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abstract | A light source having a substantially elliptical cross-section for UV curing lamp assemblies is disclosed. The light source has a pair of end sections and a central section of smaller diameter than the end sections. The end sections are each connected to the central section by a tapered section the diameter of each of which decreases from an end that mates with an end section toward an end that mates with the central section. Each of the end sections has a substantially elliptical cross-section. The central section and the tapered sections may have a substantially elliptical cross-section. The aspect ratio of the elliptical cross-section of the end sections and the central section of the light source is preferably about 2:1. |
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summary | ||
abstract | The present disclosure relates to methods and systems for calibrating an X-ray apparatus. The X-ray apparatus may include an X-ray detector and a collimator. To calibrate the X-ray apparatus, the methods and systems may include moving the X-ray detector from a first position to a second position along a first axis of a coordinate system, wherein the first position is under a scanning table, and the second position is outside the scanning table; moving the collimator to align the collimator with the X-ray detector at the second position; determining one or more parameters; and determining a second value of the first encoder when the collimator is aligned with the X-ray detector at the first position based on the one or more parameters. |
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claims | 1. A method of manufacturing a fuel element skeleton having a top nozzle, a bottom nozzle, a plurality of elongated thimble tubes having an axial dimension, the thimble tubes being supported in a spaced parallel array extending axially between the top nozzle and the bottom nozzle, a plurality of spacer grids arranged in tandem between the top nozzle and the bottom nozzle, supporting the thimble tubes in the spaced parallel array at spaced axial elevations between the top nozzle and bottom nozzle and a bottom thimble end plug extending axially from the bottom nozzle towards the top nozzle and substantially terminating at a juncture with a lowermost portion of one of the plurality of thimble tubes, wherein the bottom thimble end plug has a recess in an upper end proximate the juncture of the lowermost portion of the one of the plurality of thimble tubes, the recess in the upper end of the thimble end plug having one of either a female or male mechanically coupling contour and further including a dashpot having an outer diameter that closely approximates an interior diameter of the thimble, the dashpot including a lowermost extension with the other of either the male or female mechanically coupling contour that mates with the mechanically coupling contour in the recess in the upper end of the thimble end plug, comprising the steps of:setting up a plurality of components comprising the top nozzle, bottom nozzle, thimble tubes with the bottom thimble end plug affixed and the spacer grids in a manufacturing fixture that positions each of the components in the fuel assembly skeleton;attaching the bottom nozzle to the bottom thimble end plug;inserting a bulging tool into a top of the thimble tube;lowering the bulging tool within the thimble tube to an elevation of a lower most grid of the plurality of grids above an elevation of the thimble end plug;expanding the bulging tool to press fit the thimble tube to the lowermost grid;removing the bulging tool from the thimble tube;inserting the dashpot into the thimble tube;manipulating the dashpot to engage the female and male mechanically coupling contours in the thimble end plug recess and dashpot end plug lowermost extension;inserting the bulging tool within the thimble tool;expanding the bulging tool to press fit the dashpot to a lower section of the thimble tube and the thimble tube to the plurality of spacer grids above the lowermost grid;removing the bulging tool from the thimble tube; andremoving the components now assembled from the manufacturing fixture. 2. The method of manufacturing a fuel element skeleton of claim 1 wherein the step of attaching the bottom nozzle to the bottom thimble end plug comprises securing the bottom nozzle to the bottom thimble end plug with a standard product thimble screw. 3. The method of manufacturing a fuel element skeleton of claim 1 wherein the dashpot includes a lower dashpot end plug from which the lowermost extension extends and wherein the manipulating step comprises turning the dashpot end plug so the female and male mechanically coupling contours engage. 4. The method of manufacturing a fuel element skeleton of claim 3 wherein the dashpot end plug has an upper surface with an articulated contour and the turning step comprises engaging the articulated contour on the upper surface of the dashpot end plug with an engagement tool having a mating articulated extension that mates with the articulated contour of the upper surface on the dashpot end plug and using the engagement tool to torque the dashpot end plug in the recess in the thimble end plug. 5. The method of manufacturing a fuel element skeleton of claim 1 including the step of viewing the seating of the dashpot end plug in the recess in the bottom thimble end plug recess through a hole in a wall of the bottom thimble end plug. |
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050646079 | abstract | Hybrid nuclear reactor grey rods are described, wherein geometric combinations of relatively weak neutron absorber materials such as stainless steel, zirconium or INCONEL, and relatively strong neutron absorber materials, such as hafnium, silver-indium cadmium and boron carbide, are used to obtain the reactivity worths required to reach zero boron change load follow. One embodiment includes a grey rod which has combinations of weak and strong neutron absorber pellets in a stainless steel cladding. The respective pellets can be of differing heights. A second embodiment includes a grey rod with a relatively thick stainless steel cladding receiving relatively strong neutron absorber pellets only. A third embodiment includes annular relatively weak netron absorber pellets with a smaller diameter pellet of relatively strong absorber material contained within the aperture of each relatively weak absorber pellet. The fourth embodiment includes pellets made of a homogeneous alloy of hafnium and a relatively weak absorber material, with the percentage of hafnium chosen to obtain the desired reactivity worth. |
abstract | A system and apparatus for controlled fusion in a field reversed configuration (FRC) magnetic topology and conversion of fusion product energies directly to electric power. Preferably, plasma ions are magnetically confined in the FRC while plasma electrons are electrostatically confined in a deep energy well, created by tuning an externally applied magnetic field. In this configuration, ions and electrons may have adequate density and temperature so that upon collisions ions are fused together by the nuclear force, thus forming fusion products that emerge in the form of an annular beam. Energy is removed from the fusion product ions as they spiral past electrodes of an inverse cyclotron converter. Advantageously, the fusion fuel plasmas that can be used with the present confinement and energy conversion system include advanced (aneutronic) fuels. |
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description | The present invention relates to a spent nuclear fuel assembly storage container which stores a metal cask storing a spent nuclear fuel assembly, an assembly of the spent nuclear fuel assembly storage containers, and a method of assembling the spent nuclear fuel assembly storage container. Priority is claimed on Japanese Patent Application No. 2015-092233, filed Apr. 28, 2015, the content of which is incorporated herein by reference. Conventionally, as a storage container which safely stores radioactive contaminants, for example, storage containers disclosed in Patent Documents 1 and 2 are known. Regarding the storage container disclosed in Patent Document 1, radioactive waste is stored in a cast metal container and cast metal containers containing radioactive waste are stacked and stored. Further, regarding the radioactive contaminant storage container disclosed in Patent Documents 2, a concave portion or a convex portion is formed on each outer surface of a substantially hexagonal prism container containing radioactive contaminants within a storage space. When arranging radioactive contaminant storage containers, a convex portion of a second storage container is fitted into a concave portion of a first storage container and the outer surfaces of the substantially hexagonal prism containers are brought into contact with each other, such that they have a honeycomb structure. In this way, radioactive contaminants are stored. In contrast, unlike radioactive contaminants, spent nuclear fuel assemblies used in a reactor have a high temperature of, for example, 300° C. or more immediately after use. For that reason, spent nuclear fuel assemblies are stored in a pool of water for about 3 to 5 years and then stored in a dry storage container in a state where the ambient temperature thereof is, for example, 100° C. or less. As a dry storage container, for example, a substantially cylindrical metal cask storing a fuel assembly to which spent nuclear fuel rods are connected is known. Metal casks are in a high temperature state since nuclear fuel assemblies stored therein are reacting even when they are stored in water. For that reason, such metal casks storing nuclear fuel assemblies are stored in a strong building having earthquake resistance and a radiation shielding capability while having gaps therebetween or being laid horizontally. Further, as another dry storage means, a substantially cylindrical concrete cask having a shielding capability is known. These concrete casks storing nuclear fuel assemblies are arranged with gaps therebetween in an outdoor space. In Japan, the amount of the radiation at a site boundary needs to be limited to an allowable value of 1 mSv or less per year when storing spent nuclear fuel assemblies. Here, a storage method of storing metal casks in a building shielding radiation is selected in consideration of constraints such as the size of the site. [Patent Document 1] Japanese Unexamined Patent Application, First Publication No. 2003-167094 [Patent Document 2] Japanese Patent No. 5205540 However, due to the spent nuclear fuel assemblies reacting, the metal casks also have a high temperature. Thus, it is difficult to store the metal casks in hermetically sealed storage containers when the metal casks are in contact with each other like for the radioactive contaminants disclosed in Patent Documents 1 and 2. Further, when metal casks storing spent nuclear fuel assemblies are stored and distributed in a building having earthquake resistance and a shielding capability, costs are high. In addition, it takes costs to obtain a site for the building. The present invention has been made in view of such circumstances and an object of the present invention is to provide spent nuclear fuel assembly storage containers capable of storing spent nuclear fuel assemblies while being packed together and allowing cooling of metal casks without requiring a building having a high shielding capability, an assembly of the spent nuclear fuel assembly storage containers, and a method of assembling the spent nuclear fuel assembly storage container. Further, another object of the present invention is to provide a spent nuclear fuel assembly storage container having a sufficient shielding capability and impact resistance to improve radiation shielding performance, an assembly of the spent nuclear fuel assembly storage containers, and a method of assembling the spent nuclear fuel assembly storage container. According to a first aspect of the present invention, there is provided a spent nuclear fuel assembly storage container including: a metal cask which stores a spent nuclear fuel assembly therein; a container body which stores the metal cask and has a substantially hexagonal tubular shape; and a concave portion which is formed on an outer surface of the container body having the substantially hexagonal tubular shape, is recessed inward, and extends in a longitudinal direction, wherein the concave portion forms an external cooling passage for a cooling gas when an outer surface of the container body is joined to an outer surface of another container body. According to the first aspect of the present invention, a container body having a high radiation shielding capability and impact resistance and storing a metal cask can be stored in an outdoor place. In addition, the spent nuclear fuel assembly storage containers can store spent nuclear fuel assemblies while being packed together to form a honeycomb structure when the outer surfaces of the container bodies each having a substantially hexagonal tubular shape are in contact with each other. In addition, since the concave portion formed in any one of the outer surfaces of the container body is provided with an external cooling passage through which external air flows, the metal cask can be cooled. Therefore, an increase in temperature of the metal cask can be inhibited. According to a second aspect of the present invention, there is provided a spent nuclear fuel assembly storage container including: a metal cask which stores a spent nuclear fuel assembly; a container body which stores the metal cask and has a substantially hexagonal tubular shape; and an internal cooling passage which is provided between the metal cask and an inner surface of the container body and is provided with a cooling gas supply passage and an exhaust passage communicating with external air at upper and lower portions thereof. According to the second aspect of the present invention, the container body having high radiation shielding capability and impact resistance and storing the metal cask can be stored in an outdoor place. In addition, spent nuclear fuel assembly storage containers can store the spent nuclear fuel assemblies in an accumulation state to form a honeycomb structure while the outer surfaces of the container bodies each having a substantially hexagonal tubular shape come into contact with each other. In addition, since external air flows through the internal cooling passage between the metal cask and the inner surface of the container body, the metal cask can be cooled. Therefore, an increase in temperature of the metal cask can be suppressed. According to a third aspect of the present invention, there is provided the spent nuclear fuel assembly storage container of the first or second aspect, wherein an inner surface of the container body is provided with a stopper which prevents vibration of the metal cask. According to the third aspect of the present invention, it is possible to prevent a collision with the container body due to vibration of the metal cask by the use of a stopper even when vibration occurs due to earthquakes or the like when installing the spent nuclear fuel assembly storage container on an outdoor ground. According to a fourth aspect of the present invention, there is provided the spent nuclear fuel assembly storage container according to any one of the first to third aspects, wherein an upper surface of the container body is provided with a tapered chimney. According to the fourth aspect of the present invention, since the upper surface of the container body is provided with the tapered chimney, a gas exchanging heat with the metal cask in the external cooling passage or the internal cooling passage can be discharged from the upper surface and flows through the tapered chimney. Accordingly, since the flow speed of the gas in the external cooling passage or the internal cooling passage becomes high, the metal cask cooling effect can be promoted. According to a fifth aspect of the present invention, there is provided a spent nuclear fuel assembly storage container including: a metal cask which stores a spent nuclear fuel assembly; and a container body which stores the metal cask and has a substantially hexagonal tubular shape, wherein the container body is made of neutron shielding concrete including an aggregate mainly including colemanite and/or hilgardite collected from an ore of an evaporation type sedimentary deposit and a cement which is a consolidating material and is manufactured by mixing the cement with the colemanite and/or hilgardite as an aggregate excluding ulexite and sassolite contained in the ore of the evaporation type sedimentary deposit. According to the fifth aspect of the present invention, radiation such as neutron radiation emitted from the nuclear fuel assembly through the metal cask can be shielded by neutron shielding concrete forming the container body. For that reason, the amount of radiation emitted from the spent nuclear fuel assembly storage container to the external environment can be reduced to within allowable values. According to a sixth aspect of the present invention, there is provided the spent nuclear fuel assembly storage container according to any one of the first to fourth aspects, wherein the container body is made of heavy-weight concrete having a specific gravity of 3.5 or more. According to the sixth aspect of the present invention, radiation such as neutron rays emitted from the nuclear fuel assembly through the metal cask can be shielded by concrete having a high specific gravity and forming the container body. For that reason, the amount of the radiation emitted from the spent nuclear fuel assembly storage container to the external environment can be reduced within an allowable value. According to a seventh aspect of the present invention, there is provided an assembly of spent nuclear fuel assembly storage containers in which spent nuclear fuel assembly storage containers each including a metal cask storing a spent nuclear fuel assembly and a container body storing the metal cask and having a substantially hexagonal tubular outer surface are arranged, wherein the container bodies are arranged to have a honeycomb structure when the outer surfaces are brought into contact with each other, and wherein a space without the spent nuclear fuel assembly storage container is provided in such a manner that at least one outer surface of each of the container bodies is in contact with external air. According to the seventh aspect of the present invention, since at least one outer surface of each container body faces a space without the spent nuclear fuel assembly storage container in the spent nuclear fuel assembly storage containers arranged in an accumulation state to have a honeycomb structure, the metal cask inside the container body can be cooled by external air flowing through the space. According to an eighth aspect of the present invention, there is provided an assembly of spent nuclear fuel assembly storage containers in which spent nuclear fuel assembly storage containers each including a metal cask storing a spent nuclear fuel assembly and a container body storing the metal cask and having a substantially hexagonal tubular outer surface are arranged, wherein the container bodies are arranged to have a honeycomb structure while the outer surfaces are brought into contact with each other, and wherein container bodies not storing metal casks are arranged at a further outside of the spent nuclear fuel assembly storage containers arranged at an outside of the spent nuclear fuel assembly storage containers. According to the eighth aspect of the present invention, the radiation emitted from the metal casks inside the spent nuclear fuel assembly storage containers is shielded by the wall surfaces of the container bodies of other spent nuclear fuel assembly storage containers so that the radiation amount can be reduced within an allowable value. Part of the radiation is emitted to the outside, but is shielded while passing through the empty container bodies not storing metal casks. Thus, the radiation amount can be reduced within an allowable value. In addition, the cooling effect can be improved. According to a ninth aspect of the present invention, there is provided a method of assembling a spent nuclear fuel assembly storage container, including: installing a base; placing a metal cask storing a spent nuclear fuel assembly onto the base; and fixing the metal cask to the base by covering the metal cask using a member with a lid portion and a side cylindrical portion having a substantially hexagonal tubular shape. According to the ninth aspect of the present invention, the metal cask placed on the base is fixed while being covered by the member with the lid portion and the side cylindrical portion having a substantially hexagonal tubular shape. Accordingly, the spent nuclear fuel assembly storage container can be assembled with a small number of steps. According to the spent nuclear fuel assembly storage container of the present invention, it is possible to reduce the amount of the radiation emitted to the outside with the container body covering the metal cask and to directly or indirectly cool the metal cask stored in the container body by causing cooling gas to flow through the external cooling passage formed by the concave portion of the outer surface of the container body or the internal cooling passage formed between the metal cask and the inner surface of the container body. For that reason, even a metal cask storing a spent nuclear fuel assembly can be stored in the container body and the spent nuclear fuel assemblies can be stored while the container bodies are accumulated in such a manner that the outer surfaces of the container bodies come into contact with each other. Further, in the method of assembling the spent nuclear fuel assembly storage container, the assembly of the metal cask and the container body and the storage of the metal cask can be simply performed. Further, according to the spent nuclear fuel assembly storage container of the present invention, since the container body made of neutron shielding concrete using colemanite and/or hilgardite as an aggregate can shield radiation such as neutron rays emitted from a nuclear fuel assembly through the metal cask, the amount of the radiation emitted to the outside can be reduced. Further, according to the assembly of the spent nuclear fuel assembly storage containers of the present invention, since the spent nuclear fuel assembly storage containers are arranged to have a honeycomb structure while the outer surfaces of the substantially hexagonal tubular container bodies come into contact with each other, many storage containers can be arranged in a space occupying a small area. Further, since the amount of the radiation emitted to the outside can be reduced to an allowable value or less by the container body covering the metal cask and at least one outer surface of the spent nuclear fuel assembly storage containers arranged to have a honeycomb structure contacts external air, the metal cask inside each spent nuclear fuel assembly storage container can be cooled even in the assembly with the honeycomb structure. Further, since the amount of the radiation emitted to the outside can be reduced to an allowable value or less by the container body covering the metal cask and the empty container bodies not storing the metal casks are arranged at the further outside of the spent nuclear fuel assembly storage containers arranged at the outside of the spent nuclear fuel assembly storage containers, the amount of the radiation emitted to the outside can be reduced to an allowable value or less and the cooling effect can be improved. Hereinafter, a spent nuclear fuel assembly storage container according to a first embodiment of the present invention and an assembly thereof will be described. A spent nuclear fuel assembly storage container 1 shown in FIGS. 1 and 2 has a configuration in which a metal cask 2 storing a spent nuclear fuel assembly corresponding to an assembly of fuel rods having finished a reaction inside a nuclear reactor is stored inside a container body 3 made of concrete and having a substantially hexagonal tubular shape. The metal cask 2 storing the spent nuclear fuel assemblies has a substantially cylindrical shape. The spent nuclear fuel assembly storage container of the present invention can efficiently store the assembly of spent nuclear fuel assembly storage containers 1 in a small occupied space regardless of whether the storage place is indoors or not. In FIG. 1, the container body 3 of the spent nuclear fuel assembly storage container 1 includes a hexagonal plate-shaped base 5 on which the metal cask 2 is placed; a side cylindrical portion 6 of which an outer surface 6a is formed in a substantially hexagonal tubular shape and an inner surface 6b is formed in a substantially cylindrical shape, and a substantially disc-shaped lid portion 7 which is provided at an upper opening portion of the side cylindrical portion 6. The lid portion 7 is fitted to the upper opening portion 6c of the side cylindrical portion 6 and is integrally held as a hexagonal tube lid. The lid portion 7 can be separated from the side cylindrical portion 6 if necessary. All of the base 5, the side cylindrical portion 6, and the lid portion 7 are made of concrete capable of shielding radiation such as neutrons emitted from the spent nuclear fuel assemblies. The container body 3 may be made of a material other than concrete as long as a radiation shielding capability can be exhibited. In the side cylindrical portion 6 of the container body 3 shown in FIG. 2, the upper portion of the cylindrical inner surface 6b storing the metal cask 2 forms the stepped opening portion 6c opening upward in which the upper portion of the cylindrical inner surface 6b widens its diameter to serve a circular stepped shape. The disc-shaped lid portion 7 having a stepped outer peripheral surface is fitted to the opening portion 6c and is fixed to the upper portion of the side cylindrical portion 6 at an enlarged disc portion 7a of the lid portion 7 with bolts or the like. In addition, a joint portion between the base 5 and the side cylindrical portion 6 of the container body 3 also has a stepped shape similar to the fitting portion between the lid portion 7 and the opening portion 6c of the side cylindrical portion 6. Accordingly, radiation emitted from the metal cask 2 can be reliably shielded by the side cylindrical portion 6. The side cylindrical portion 6 of the container body 3 includes six outer surfaces 6a which is similar to a rectangular hexagonal tube and each outer surface 6a includes a concave portion 10 formed at a center portion in the width direction. The concave portion 10 is formed along the entire length in the longitudinal direction from the lower end to the upper end of the center of the outer surface 6a and is formed to extend to each surface of the base 5. The concave portion 10 forms an external cooling passage 10A which is formed in the outer surface 6a of the container body 3 and through which external air passes. When external air flows along the concave portion 10 between the upper and lower sides in each outer surface 6a, the metal cask 2 can be cooled through the container body 3. Further, the inner surface 6b of the side cylindrical portion 6 of the container body 3 is formed in a substantially cylindrical shape with a gap between the inner surface 6b and the metal cask 2 being formed. An air supply passage 13 provided with an air supply port communicating with the outer surface 6a of the side cylindrical portion 6 is formed in the vicinity of the base 5 at the lower portion of the side cylindrical portion 6 to extend in the radial direction. The air supply passage 13 communicates with the lower end portion of the substantially cylindrical internal cooling passage 12. In addition, each air supply passage 13 provided in the side cylindrical portion 6 is bent radially in a stepped shape from the internal cooling passage 12, extends outward in the radial direction, and opens to the concave portion 10 of one of the six outer surfaces 6a. Accordingly, the radiation emitted from the metal cask 2 into the air supply passage 13 can be shielded by the wall surface of the side cylindrical portion 6. Further, the upper portion of the internal cooling passage 12 in the side cylindrical portion 6 communicates with an exhaust passage 14 provided with an exhaust port which is formed to have a substantially L-shape in the cross-section inside the side cylindrical portion 6 and opens to the upper surface 6d. In addition, each exhaust passage 14 radially extends outward in the radial direction from the vicinity of the upper end of the internal cooling passage 12 and opens to the vicinity of one of the corner portions of the upper surface 6d formed in a hexagonal shape. The exhaust port of the exhaust passage 14 provided in the upper surface 6d may be located immediately above the air supply passage 13 and may be formed at an arbitrary position in the upper surface 6d. However, in this embodiment, the exhaust port of the exhaust passage 14 is formed in a thick corner portion. Further, in the internal cooling passage 12 inside the container body 3 shown in FIG. 2, stoppers 15 are attached to the inner surface 6b via an attachment member 16 in order to inhibit vibration of the metal cask 2 inside a gap K between the inner surface 6b and the metal cask 2 due to earthquakes or the like. As shown in FIG. 3, each of the stoppers 15 is preferably made of, for example, metal withstanding a high temperature and forms, for example, a minute gap of about 10 to 15 mm between each of the stoppers 15 and the metal cask 2. In order to prevent a problem in which the metal cask 2 is vibrated and collides with the inner surface 6b of the side cylindrical portion 6, the stoppers 15 are preferably provided at predetermined intervals, for example, intervals of 90° in the circumferential direction on the inner surface 6b of the side cylindrical portion 6 and also have appropriate gaps therebetween in the vertical direction. Alternatively, the stoppers 15 may be arranged in a spiral shape along the inner surface 6b. The spent nuclear fuel assembly storage container 1 according to this embodiment has the above-described configuration. Next, a method of assembling the spent nuclear fuel assembly storage container 1 will be described with reference to FIGS. 1 and 2. Since the spent nuclear fuel assembly storage container 1 according to this embodiment is installed on, for example, an outdoor ground, the base 5 is provided on the ground. Then, the metal cask 2 is placed on the center of the base 5. Next, when the metal cask 2 is covered with the side cylindrical portion 6 having the lid portion 7 fitted thereto from above the metal cask 2, the metal cask 2 is surrounded by the container body 3 including the base 5, the side cylindrical portion 6, and the lid portion 7. Finally, when the side cylindrical portion 6 and the base 5 are fixed to each other with screws, the assembly of the spent nuclear fuel assembly storage container 1 is completed. An operation of an assembly 18 using the spent nuclear fuel assembly storage containers 1 obtained in this way will be described. In the spent nuclear fuel assembly storage containers 1 according to this embodiment, the container bodies 3 are densely arranged while the outer surfaces 6a of the container bodies 3 are respectively brought into contact with each other. Therefore, the assembly 18 having a honeycomb structure can be formed as shown in FIG. 4. As shown in FIG. 5, in the assembly 18 of the spent nuclear fuel assembly storage containers 1, a first spent nuclear fuel assembly storage container 1 and a second spent nuclear fuel assembly storage container 1 come into contact with each other at the outer surfaces 6a to form the external cooling passage 10A using the pair of concave portions 10 provided in each outer surface 6a. In addition, the air supply passage 13 communicating with the internal cooling passage 12 opens to the concave portion 10 of each outer surface 6a. For this reason, even in the state of the assembly 18 in which spent nuclear fuel assembly storage containers 1 are assembled such that they are in contact with each other in a honeycomb structure on the outdoor ground, external air flows from the top to the bottom inside the external cooling passage 10A formed by the pair of concave portions 10 of the adhering outer surfaces 6a. Then, the external air passes through the air supply passage 13 formed below the concave portion 10 of each outer surface 6a of each spent nuclear fuel assembly storage container 1 and flows inside the internal cooling passage 12 between the metal cask 2 and the inner surface 6b of the container body 3 to cool the metal cask 2. Then, the high-temperature external air obtained by heat exchange with the metal cask 2 passes through the exhaust passage 14 communicating with the upper surface 6d of each spent nuclear fuel assembly storage container 1 to be discharged to the outside. Thus, the external cooling passage 10A provided in each of the six outer surfaces 6a of each spent nuclear fuel assembly storage container 1 and the cooling passage passing through the air supply passage 13, the internal cooling passage 12, and the exhaust passage 14 are formed even when the spent nuclear fuel assembly storage containers 1 are brought into contact with each other to form a honeycomb structure. Therefore, the metal cask 2 inside the container body 3 can be efficiently cooled. In addition, the radiation emitted through the metal cask 2 stored inside the spent nuclear fuel assembly storage container 1 is attenuated by the concrete container body 3 surrounding the metal cask 2 and is further attenuated by the container bodies 3 of the spent nuclear fuel assembly storage containers 1 adjacent to the above spent nuclear fuel assembly storage container 1. For this reason, the amount of radiation emitted to the external air from the outer spent nuclear fuel assembly storage containers 1 of the assembly 18 can be minimized to be less than an allowable range of 1 mSv or less per year. Further, as shown in FIG. 4 and (a) of FIG. 6, in this embodiment, the assembly 18 of the spent nuclear fuel assembly storage containers 1 is stored while they are arranged in a close contact state having a honeycomb structure in an outdoor site. For this reason, it is possible to prevent falling down of the spent nuclear fuel assembly storage containers 1 due to interference therebetween even in the event of earthquakes. As shown in (b) of FIG. 6, the assembly 18 of the spent nuclear fuel assembly storage containers 1 may be holded in an outdoor site while being restrained by a wire 17. As shown in (c) of FIG. 6, the assembly 18 of the spent nuclear fuel assembly storage containers 1 may be holded while being placed on or a flat dish-shaped member 19, a surface of a floor, or the like. In such a case, each of the spent nuclear fuel assembly storage containers 1 of the assembly 18 and the metal casks 2 stored therein do not collapse or become damaged. Due to this, the spent nuclear fuel assembly storage container 1 according to this embodiment or the assembly 18 thereof does not need be stored in a shielded building with earthquake resistance. In addition, a rigid foundation on which a container or an assembly thereof may be placed does not need to be provided. For that reason, the spent nuclear fuel assembly storage container 1 can be easily stored and transported. Further, the spent nuclear fuel assembly storage container 1 can store the spent nuclear fuel assemblies in such a manner that the spent nuclear fuel assembly storage containers 1 are arranged to be brought into contact with each other in the horizontal direction and to overlap each other in the upward direction. As described above, according to the assembly 18 of the spent nuclear fuel assembly storage containers 1 of this embodiment, the spent nuclear fuel assembly storage container 1 can cool the metal cask 2 using the external cooling passage 10A formed by the concave portion 10 provided in each substantially hexagonal outer surface 6a, the air supply passage 13 provided in the concave portion 10, the internal cooling passage 12, and the exhaust passage 14. For that reason, even when the spent nuclear fuel assembly storage containers 1 are stored as the assembly 18 to have a honeycomb structure in a close contact state, all the metal casks 2 can be cooled using the external cooling passages 10A formed by the concave portions 10 of the spent nuclear fuel assembly storage containers 1. In addition, since the radiation emitted from the metal cask 2 is shielded by the concrete container body 3, the amount of the radiation emitted through the metal cask 2 can be attenuated. Further, since the radiation passing through the container body 3 is also shielded by the container bodies 3 of other adjacent spent nuclear fuel assembly storage containers 1, the amount of the radiation can be further attenuated. For that reason, since the annual radiation amount at a boundary between the assembly 18 and the external environment thereof can be minimized to an allowable value of 1 mSv or less, safety is ensured. Further, since the spent nuclear fuel assembly storage containers 1 do not need to be stored while being separated from each other within a building which shields the radiation, a strong foundation ground does not need to be provided, and the spent nuclear fuel assemblies are arranged on an outdoor ground in a close contact state, the spent nuclear fuel assemblies can be easily stored and transported. Therefore, since the space occupied by the spent nuclear fuel assemblies is small, the spent nuclear fuel assemblies can be stored and transported at low cost. In addition, the spent nuclear fuel assembly storage container 1 according to the present invention and the assembly 18 thereof are not limited to those of the above-described embodiment, and various modifications or replacements can be made without departing from the spirit of the present invention. Hereinafter, other embodiments or modified examples of the present invention will be described, but the same reference numerals will be used for components and members the same as or similar to those of the above-described embodiment. Next, FIG. 7 shows an assembly 20 of the spent nuclear fuel assembly storage containers 1 according to a second embodiment of the present invention. In FIG. 7, since the spent nuclear fuel assembly storage containers 1 are arranged in a close contact state to have a honeycomb structure, the radiation emitted from the metal cask 2 passing through the container body 3 of each of the spent nuclear fuel assembly storage containers 1 is shielded and attenuated while passing through the wall surfaces of the side cylindrical portions 6. For that reason, the radiation amount at a boundary between the assembly 20 and the external environment thereof can be reliably within an allowable value of 1 mSv or less per year. However, in a spent nuclear fuel assembly storage container 1 disposed at the outermost side of the assembly 20, the radiation emitted from the metal cask 2 is discharged to the external environment while passing through only the wall surface of one side cylindrical portion 6. For this reason, there is a possibility that the radiation amount may be larger than the annual allowable value. For that reason, in the second embodiment, dummy spent nuclear fuel assembly storage containers 1A including empty container bodies 3 not storing the metal casks 2 are arranged at the outermost side of the assembly 20 of the spent nuclear fuel assembly storage containers 1 so that the spent nuclear fuel assembly storage containers 1 are surrounded by the dummy spent nuclear fuel assembly storage containers 1A. By employing such a configuration, the radiation emitted outward from a spent nuclear fuel assembly storage container 1 disposed at the outermost side in the assembly 20 of the spent nuclear fuel assembly storage containers 1 is shielded while passing through the wall surfaces of the side cylindrical portions 6 of the dummy spent nuclear fuel assembly storage containers 1A. Therefore, the annual radiation amount may be able to be brought within the allowable value of 1 mSv or less. In addition, the spent nuclear fuel assembly storage containers 1 which are disposed inside the empty spent nuclear fuel assembly storage containers 1A can be cooled by the empty spent nuclear fuel assembly storage containers 1A. In addition, when the radiation amount is not sufficiently lowered even by a configuration in which a single row of dummy spent nuclear fuel assembly storage containers 1A are arranged at the outside of the assembly 20 to surround the spent nuclear fuel assembly storage containers 1, for example, double, triple, or multiple rows of dummy spent nuclear fuel assembly storage containers 1A may be arranged at the outside so that the annual radiation amount becomes reliably within the allowable value of 1 mSv or less. Next, FIG. 8 shows an assembly 22 of the spent nuclear fuel assembly storage containers 1 according to a third embodiment of the present invention. In the assembly 22 of the spent nuclear fuel assembly storage containers 1 shown in FIG. 8, similarly to the second embodiment, the spent nuclear fuel assembly storage containers 1 are arranged to form a honeycomb structure in a close contact state and the outside thereof is surrounded by the dummy spent nuclear fuel assembly storage containers 1A. However, substantially hexagonal tubular spaces in which the inner spent nuclear fuel assembly storage containers 1 are removed are formed at predetermined intervals. These spaces will be referred to as void holes 23. In this embodiment, the void holes 23 are arranged in such a manner that at least one outer surface 6a of each one of the substantially hexagonal tubular container bodies 3 faces the void hole 23 in spent nuclear fuel assembly storage containers 1 other than the outer spent nuclear fuel assembly storage containers 1. Accordingly, in the spent nuclear fuel assembly storage containers 1 except for the dummy spent nuclear fuel assembly storage containers 1A, when external air passes through the external cooling passage 10A formed by the concave portion 10 in the six outer surfaces 6a and the internal cooling passage 12 inside the spent nuclear fuel assembly storage container 1 and flows to come into contact with the metal cask 2, the metal cask 2 can be cooled. In addition, since the external cooling passage 10A formed by the concave portion 10 communicates with the void hole 23 in six spent nuclear fuel assembly storage containers 1 facing the void hole 23, the function of cooling the metal cask 2 is further improved. In addition, when the dummy spent nuclear fuel assembly storage containers 1A are not provided at the outside of the assembly 22 of the spent storage containers 1, the effect of cooling the metal cask 2 in the outer spent nuclear fuel assembly storage containers 1 is further improved. Further, as a modified example of the third embodiment, the dummy spent nuclear fuel assembly storage containers 1A in the assembly 22 of the spent nuclear fuel assembly storage containers 1 may be removed, and spent nuclear fuel assembly storage containers 24 shown in FIG. 9 and including container bodies 3 without concave portions 10 may be provided instead of other spent nuclear fuel assembly storage containers 1. In this case, external air falling into the void hole 23 enters the internal cooling passage 12 from the air supply passage 13 with the air supply port provided in the outer surface 6a facing the void hole 23 to cool the metal cask 2, and passes through the exhaust passage 14 of the upper surface 6d to be discharged to the atmosphere. Thus, according to the assembly 22 of this modified example, the metal cask 2 inside the spent nuclear fuel assembly storage container 24 can be cooled without forming the concave portion 10 in the outer surface 6a of the container body 3. Next, a spent nuclear fuel assembly storage container 26 according to a fourth embodiment of the present invention will be described with reference to FIG. 10. A basic configuration of the spent nuclear fuel assembly storage container 26 according to this embodiment is the same as that of the spent nuclear fuel assembly storage container 1 according to the first embodiment. However, the spent nuclear fuel assembly storage container 26 has a difference in that the exhaust passage 14 extends along a line extending from the internal cooling passage 12 and opens to the upper surface of the lid portion 7. Further, a chimney 28 which is a cooling tower is attached to the upper surface 6d of the side cylindrical portion 6 of the container body 3. The chimney 28 includes a small-diameter opening portion (outlet) formed at an upper end of a roof portion 28a which decreases in diameter upward from a large-diameter opening portion (inlet) covering the upper surface 6d of the container body 3 to formed a tapered shape, and a second roof portion 28b formed at the upper surface. For that reason, when an exhaust gas discharged from the opening portion of the exhaust passage 14 of the container body 3 is caused to converge by the chimney 28, the flow speed of the exhaust gas increases so that the cooling efficiency of the metal cask 2 is improved. In addition, the chimney 28 is made of steel, but may be made of concrete. Further, the chimney 28 may be integrated with or separate from the container body 3. According to the spent nuclear fuel assembly storage container 26 of the fourth embodiment with the above-described configuration, since external air flowing into the air supply passage 13 from the air supply port of the container body 3 flows through the internal cooling passage 12 from the bottom toward the top, the outer peripheral surface of the metal cask 2 can be efficiently cooled. In addition, the chimney 28 is provided in the upper surface 6d of the container body 3 so that the external air is discharged from the exhaust port of the exhaust passage 14 to the chimney 28 at a high flow speed. Therefore, the flow speed of the internal cooling passage 12 is further increased and thus the effect of cooling the metal cask 2 can be further improved. Further, concrete capable of shielding radiation is used as a material of the container body 3 storing the metal cask 2 by the spent nuclear fuel assembly storage container 1. As concrete capable of shielding radiation, various types of concrete including ordinary concrete using Portland cement and the like can be used. In particular, it is preferable to use concrete having a high neutron shielding performance among radiation shielding performances for the container body 3. In this embodiment, the neutron shielding concrete used as the material of the container body 3 will be described below. The neutron shielding concrete according to this embodiment includes a boron-based boron aggregate such as colemanite and/or hilgardite selected from ores of an evaporation type sedimentary deposit and cement which is a consolidating material and is manufactured by mixing the cement with colemanite and/or hilgardite corresponding to the aggregate except for eurekite and sassolite contained in the ores of the evaporation type sedimentary deposit. Further, since ulexite and sassolite contained in ores of an evaporation type sedimentary deposit easily dissolve in water and may dissolve in water beforehand to inhibit the hydration reaction of cement, colemanite and/or hilgardite may not be easily solidified with the cement. However, according to the above-described neutron shielding concrete, when colemanite or hilgardite which is a boron mineral selected and collected from the ores of the evaporation type sedimentary deposit is kneaded and mixed with cement except for eurekite and sassolite contained in the ores of the evaporation type deposit, it is possible to obtain a boron-containing concrete having high strength, high durability, and sufficient neutron shielding performance. This neutron shielding concrete is disclosed in detail in Japanese Patent Application No. 2013-272471 proposed by the present inventor. Further, as other kinds of concrete having high radiation shielding performance, concrete (G-concrete) having high specific gravity of 3.5 or more can be used. This high specific gravity concrete mainly includes heavy aggregates such as sand, gravel, magnetite, iron ore, iron oxide powder, and oxidized slag. In the spent nuclear fuel assembly storage container 1 according to this embodiment, the container body 3 is made of the above-described boron-containing concrete including boron minerals or high specific gravity concrete having a specific gravity of 3.5 or more or is made by applying or laminating boron-containing concrete or high specific gravity concrete to an outer or inner surface of ordinary concrete. When such concrete is used as the material of the container body 3, it is possible to reduce the transmission amount of the radiation emitted from the metal cask 2 by shielding the radiation and to minimize the radiation amount at the boundary between the assembly 18 and the external environment thereof to 1 mSv or less per year. In addition, in the above-described first embodiment, the side cylindrical portion 6 of the spent nuclear fuel assembly storage container 1 is formed in a hexagonal tubular shape and each outer surface 6a is provided with the concave portion 10. Accordingly, when adjacent spent nuclear fuel assembly storage containers 1 are brought into contact with each other, the external cooling passage 10A having a substantially hexagonal tubular shape is formed between the concave portions 10. However, instead of this configuration, when the concave portion 10 is provided on one outer surface 6a of the spent nuclear fuel assembly storage containers 1 which come into contact with each other as shown in FIG. 11, the external cooling passage 10A communicating with the air supply passage 13 can be formed. Further, the air supply passage 13 and the exhaust passage 14 communicating with the internal cooling passage 12 or the external cooling passage 10A may not be formed in all six surfaces of the container body 3 and may be formed in at least one of six surfaces. Further, in the present invention, the spent nuclear fuel assembly storage container 1 may be provided with any one of the external cooling passage 10A formed by the concave portion 10 of the outer surface 6a of the container body 3 and the internal cooling passage 12 formed between the metal cask 2 and the inner surface 6b of the container body 3. When the external cooling passage 10A is provided by the concave portion 10, the metal cask 2 can be indirectly cooled through the container body 3 by the external air flowing in the concave portion 10. Further, when the chimney 28 is provided in the upper surface 6d of the container body 3, the circulation of external air in the external cooling passage 10A flows faster. Therefore, the cooling effect can be improved. Further, when the internal cooling passage 12 is provided, the metal cask 2 can be directly cooled by the external air flowing through the air supply passage 13 and the exhaust passage 14. According to the assembly of the spent nuclear fuel assembly storage containers of the present invention, a large number of storage containers can be disposed in a small occupied space, the amount of the radiation emitted to the outside can be reduced to the allowable value or less, and the cooling effect can be improved. 1, 24, 26 Spent nuclear fuel assembly storage container 1A Dummy spent nuclear fuel assembly storage container 2 Metal cask 3 Container body 5 Base 6 Side cylindrical portion 6a Outer surface 6b Inner surface 6d Upper surface 10 Concave portion (external cooling passage) 10A External cooling passage 12 Internal cooling passage 13 Air supply passage 14 Exhaust passage 15 Stopper 18, 20 Assembly 23 Void hole 27 Fin 28 Chimney |
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048093149 | abstract | An X-ray inspection system includes an X-ray source for generating a directed X-ray beam and a linear array detector for measuring the intensity of the received radiation and generating electrical signals representative thereof. A method for aligning the detector with the directed X-ray beam includes removing any part between the X-ray source and the detector, opening an X-ray beam limiter, positioning the linear array detector for maximum signal from each detector element, reducing the X-ray beam limiter opening, detecting whether any signal from a detector element is reduced, moving the limiter for producing a maximum signal on each signal, securing the X-ray beam limiter, and positioning the detector array for maximum signal. |
047553502 | claims | 1. A thermionic energy conversion system assembly, comprising: (a) a source of heat; (b) a plurality of emitter electrodes thermally coupled to and generally enclosed by the source of heat; (c) a plurality of collector electrodes in a spaced relationship from and generally enclosed by the emitter electrodes; and, (d) a heat sink generally enclosed by the collector electrodes. (a) a source of heat; (b) a plurality of emitter electrodes thermally coupled to and generally enclosed by the source of heat; (c) a plurality of collector electrodes in a spaced relationship from and generally enclosed by the emitter electrodes; (d) a heat sink generally enclosed by the collector electrodes; (e) wherein the heat sink comprises a container of heat sink material; and, (f) wherein the heat sink material comprises a lithium salt. (a) a shource of heat; (b) a plurality of emitter electrodes thermally coupled to and generally enclosed by the source of heat; (c) a plurality of collector electrodes in a spaced relationship from and generally enclosed by the emitter electrodes; (d) a heat sink generally enclosed by the collector electrodes; (e) wherein the heat sink comprises a container of heat sink material; and, (f) wherein the heat sink material comprises a neutron moderator. (a) a source of heat; (b) a plurality of emitter electrodes thermally coupled to and generally enclosed by the source of heat; (c) a plurality of collector electrodes in a spaced relationship from and generally enclosed by the emitter electrodes; and, (d) a heat sink generally enclosed by the collector electrodes. (a) a source of heat; (b) a plurality of emitter electrodes thermally coupled to and generally enclosed by the source of heat; (c) a plurality of collector electrodes in a spaced relationship from and generally enclosed by the emitter electrodes; and, (d) a heat sink generally enclosed by the collector electrodes. (a) providing a source of heat; (b) thermally coupling the heat source to a plurality of emitter electrodes generally enclosed by the source of heat; (c) providing a plurality of collector electrodes in a spaced relationship from and generally enclosed by the emitter electrodes; and (d) thermally coupling the collector electrodes to a heat sink generally enclosed by the collector electrodes. 2. A thermionic energy conversion system assembly according to claim 1, wherein the source of heat comprises fissionable material. 3. A thermionic energy conversion system assembly according to claim 1, wherein the heat sink comprises a container of heat sink material. 4. A thermionic energy conversion system assembly comprising: 5. A thermionic energy conversion system assembly according to claim 4, wherein the lithium salt comprises lithium hydride. 6. A thermionic energy conversion system assembly comprising: 7. A thermionic energy conversion system assembly according to claim 6, wherein the heat sink material comprises lithium hydride enriched in the Li-7 isotope. 8. A thermionic energy conversion system assembly according to claim 3, further comprising a heat pipe, enclosed generally within the heat sink material, for transferring heat out of the heat sink material. 9. A thermionic energy conversion system module, comprising a plurality of stacked thermionic conversion system assemblies, each assembly comprising: 10. A thermionic energy conversion system module according to claim 9, wherein the heat sinks are generally continuous from assembly to assembly. 11. A thermionic energy conversion system module according to claim 10, wherein the heat sinks comprise a container of heat sink material, and further comprise a heat pipe, enclosed generally within the heat sink material, for transferring heat out of the heat sink material. 12. A thermionic nuclear reactor, comprising an array of thermionic energy conversion system modules, each module comprising a plurality of stacked thermionic conversion system assemblies, each assembly comprising: 13. A thermionic nuclear reactor according to claim 12, wherein each heat sink comprises a container of heat sink material, and each module further comprises a heat pipe, enclosed generally within the heat sink material, for transferring heat out of the heat sink material. 14. A method of converting heat energy to electrical energy, comprising the steps of: 15. The method of converting heat energy to electrical energy according to claim 14, further comprising the step of removing heat from the heat sink with a heat pipe. |
claims | 1. A laser apparatus comprising:a gas spraying portion comprising a slit and a plurality of opening portions;a laser oscillator which is provided over the gas spraying portion and outside the slit and is capable of outputting laser light;an optical system being arranged to allow the laser light to penetrate the slit and to irradiate a process object;a rotating base which is capable of placing a stage thereon and which moves along a straight line when the process object is irradiated with the laser light; andmeans for rotating the rotating base when the process object is irradiated with the laser light,wherein the gas spraying portion is arranged to allow a gas to be sprayed through the plurality of opening portions on the process object when the process object is irradiated with the laser light. 2. The laser apparatus according to claim 1,wherein the optical system is capable of processing the laser light into a linear shape. 3. The laser apparatus according to claim 1,wherein the gas spraying portion is placed over the rotating base. 4. The laser apparatus according to claim 1,wherein the laser oscillator is a continuous wave solid-state laser. 5. The laser apparatus according to claim 1,wherein the laser oscillator is selected from a continuous wave YAG laser, YVO4 laser, YLF laser, YAlO3 laser, glass laser, ruby laser, alexandrite laser, Ti: sapphire laser, and Y2O3 laser. 6. The laser apparatus according to claim 1,wherein the laser oscillator is selected from a continuous wave excimer laser, Ar laser, and Kr laser. 7. The laser apparatus according to claim 1,wherein the laser light is a second harmonic. 8. A laser apparatus comprising:a gas spraying portion comprising a slit and a plurality of opening portions;a laser oscillator which is provided over the gas spraying portion and outside the slit and is capable of outputting laser light;an optical system being arranged to allow the laser light to penetrate the slit and to irradiate a process object;a rotating base which is capable of placing a plurality of stages thereon and which moves along a straight line when the process object is irradiated with the laser light; andmeans for rotating the rotating base when the process object is irradiated with the laser light,wherein the gas spraying portion is arranged to allow a gas to be sprayed through the plurality of opening portions on the process object when the process object is irradiated with the laser light. 9. The laser apparatus according to claim 8,wherein the optical system is capable of processing the laser light into a linear shape. 10. The laser apparatus according to claim 8,wherein the gas spraying portion is placed over the rotating base. 11. The laser apparatus according to claim 8,wherein the laser oscillator is a continuous wave solid-state laser. 12. The laser apparatus according to claim 8,wherein the laser oscillator is selected from a continuous wave YAG laser, YVO4 laser, YLF laser, YAlO3 laser, glass laser, ruby laser, alexandrite laser, Ti: sapphire laser, and Y2O3 laser. 13. The laser apparatus according to claim 8,wherein the laser oscillator is selected from a continuous wave excimer laser, Ar laser, and Kr laser. 14. The laser apparatus according to claim 8,wherein the laser light is second harmonic. 15. The laser apparatus according to claim 1,wherein the slit is stationarily surrounded by the plurality of opening portions. 16. The laser apparatus according to claim 8,wherein the slit is stationarily surrounded by the plurality of opening portions. 17. The laser apparatus according to claim 1,wherein the laser apparatus is arranged to allow the process object to be irradiated a plurality of times, andwherein a moving distance of the rotating base along the straight line during one rotation thereof is smaller than a width of the laser light. 18. The laser apparatus according to claim 8,wherein the laser apparatus is arranged to allow the process object to be irradiated a plurality of times, andwherein a moving distance of the rotating base along the straight line during one rotation thereof is smaller than a width of the laser light. 19. The laser apparatus according to claim 1,wherein a reflector is provided over the stage. 20. The laser apparatus according to claim 8,wherein a reflector is provided over each of the plurality of stages. 21. The laser apparatus according to claim 1,wherein a reflector is provided over the stage, andwherein a surface of the reflector is waved. 22. The laser apparatus according to claim 8,wherein a reflector is provided over each of the plurality of stages, andwherein a surface of the reflector is waved. 23. The laser apparatus according to claim 1,wherein the slit, the plurality of opening portions, and a conduit through which the gas is supplied to the gas spraying portion are incorporated with the gas spraying portion. 24. The laser apparatus according to claim 8,wherein the slit, the plurality of opening portions, and a conduit through which the gas is supplied to the gas spraying portion are incorporated with the gas spraying portion. |
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description | The present invention concerns a method of making a nuclear fuel pellet for a nuclear power reactor. Different manners of producing nuclear fuel pellets are known by a person skilled in the art. It is common to make the nuclear fuel pellet from a nuclear fuel material in powder form. The nuclear fuel material may for example be UO2, where U is enriched with regard to 235U. The powder material may also include other additives, such as U3O8 and binder material. The powder is pressed in order to form a so-called green pellet. The concept “green pellet” in this technical field means the pressed pellet before it is sintered. The green pellet is thus thereafter sintered in a furnace. The sintered pellets are thereafter ground in order to obtain the correct diameter and surface finish. It is also known to include some additives in the powder in order to increase the grain size in the sintered pellet. For example WO 00/49621 A1 gives some examples of such additives and describes how the nuclear fuel pellet may be produced. DE 3235944 A1 describes that a solution of hydrogen peroxide is added to the uranium dioxide powder in order to increase the grain size. WO 2005/041208 A2 describes that a porous uranium dioxide arrangement is in-filtrated with a precursor liquid in the form of allylhydridopolycarbosilane in order to enhance the thermal conductivity in the nuclear fuel. As a background of the present invention a further phenomenon should be mentioned. This phenomenon is a structure in the used nuclear fuel pellets called high burn-up structure (HBS) or rim structure. When the nuclear fuel has been used for a longer time in a nuclear reactor (i.e. a high burn-up) a new restructured configuration appears at the outer thin region of the fuel pellet. This phenomenon is described for example in the article “The high burn-up structure in nuclear fuel” by V. V. Rondinella et al. in Materials Today, December 2010, Volume 13, No. 12, pages 24-32. The HBS means that the grains in the outer region of the nuclear fuel pellet subdivide into very small grains. The outer region in which the HBS appears may for example be less than 100 μm thick. When in this document a certain percentage of a material is mentioned, this concerns weight percent, if nothing else is said. When in this document a certain grain size is mentioned, this refers to the so-called two dimensional (2D) grain size, i.e. the grain size measured in a plane, if nothing else is said. An object of the present invention is to provide a method of making a nuclear fuel pellet, with which method it is possible to better control the addition of an additive to the nuclear fuel pellet, which additive increases the grain size. A further object is to provide such a method which can be carried out in a relatively simple manner. At least the above objects are achieved by a method of making a nuclear fuel pellet for a nuclear power reactor, the method comprising the following steps: providing a nuclear fuel material in powder form; pressing the powder such that a so-called green pellet is obtained; providing a liquid that comprises an additive which is to be added to the green pellet; contacting the green pellet with the liquid such that the liquid, with the additive, penetrates into the pellet; and sintering the so treated green pellet, wherein said additive is such that larger grains in the nuclear fuel material are present in the pellet after the sintering step as compared with the grain size obtained if a pellet to which no such additive has been added is sintered in the same manner. According to the invention, the additive which increases the grain size is thus added after the green pellet has been formed. It is therefore not necessary to add the additive to the powder before pressing the green pellet. Since the additive is provided in a liquid, it can be controlled to which extent the additive enters into the green pellet. An improved control of the addition of the additive which increases the grain size is therefore achieved. Furthermore, it is quite easy to apply the liquid, with additive, to the green pellet. According to one manner of carrying out the method according to the invention, said additive is in the form of particles dispersed in said liquid. The additive is thus in the form of particles which may penetrate into the pores in the green pellet. It is advantageous to use a liquid as a carrier of such particles. Since the particles are dispersed in the liquid, the particles do not dissolve in the liquid. According to a further manner of carrying out the method according to the invention, the method comprises a step of controlling the penetration depth of the liquid, and thereby of the additive, into the green pellet. By controlling the penetration depth, it is possible to control in which region in the pellet the liquid, with the additive, is present. It is thereby possible to control where the additive is present in the pellet. According to a further manner of carrying out the method according to the invention, said step of controlling the penetration depth is done by selecting one or both of the following: the viscosity of the liquid with included additive, the amount of the liquid, with the additive, which is added to the green pellet when contacting the green pellet with the liquid, with the additive. By selecting a liquid with a certain viscosity it is possible to control the penetration depth of the liquid. The penetration depth may also be controlled by controlling how much liquid is added to the green pellet. The penetration depth can also be controlled by selecting a certain particle size for the additive particles or by selecting a certain particle size distribution. Smaller particles tend to penetrate deeper into the green pellet than larger particles. The amount of the liquid, with the additive, which is added to the green pellet can be controlled for example by spraying a certain amount of the liquid, with additive, onto the green pellet, or by exposing the green pellet to the liquid, with additive (for example by dipping the green pellet in the liquid, with additive) during a predetermined time. According to a further manner of carrying out the method according to the invention, the penetration depth of the liquid, with the additive, into the green pellet is controlled such that an outer portion of the green pellet contains substantially more liquid, and thereby more additive, than an inner portion of the green pellet, such that the sintered pellet has a larger grain size in the outer portion than in the inner portion. It is thus possible to control the grain size to be larger in an outer portion of the sintered pellet. The inventors of the present invention have realized that the occurrence of the above described HBS may be prevented, or delayed, if the nuclear fuel pellet has larger grains in the outer portion of the nuclear fuel pellet (where the HBS occurs). The outer and inner portions may be defined in different manners. For example, if we consider a cylindrical nuclear fuel pellet with a radius r, the inner portion may for example be the part of the nuclear fuel pellet from the centre of the pellet outwards up to for example 0.6 r and the outer portion may for example be the part of the nuclear fuel pellet that is located between 0.8 r and r or between 0.9 r and r, or between 0.95 r and r (depending on where it is desired that the grains are larger). When it is stated that the grain size is larger in the outer portion, also this may be defined in different manners. For example, if we consider the average 2D grain size in the outer portion and the average 2D grain size in the inner portion, the average grain size in the outer portion may be at least 50%, preferably at least 100%, larger than the average grain size in the inner portion. According to a further manner of carrying out the method according to the invention, said liquid with additive is selected and said method is performed such that the liquid with additive will penetrate into the pores which exist between the grains in the green pellet. The green pellet will have pores both between the grains in the green pellet and inside the grains in the green pellet. The pores inside the grains are normally smaller than the pores which exist between the grains. Consequently, it can be controlled (for example by selecting a certain viscosity) that the liquid will penetrate into the pores which exist between the grains. According to a further manner of carrying out the method according to the invention, said liquid with additive is selected and said method is performed such that the liquid with additive will not, at least not to any substantial degree, penetrate into the pores which exist in the grains in the green pellet. According to this alternative, the additive will not to any substantial degree enter into the grains, but the additive will be added into the pores which exist between the grains. According to a further manner of carrying out the method according to the invention, said liquid with additive is selected and said method is performed such that the liquid with additive will penetrate also into the pores which exist in the grains in the green pellet. According to this alternative, the additive will thus enter also into the pores in the grains. With the present invention it is thus possible to control where in the green pellet the additive is added. According to a further manner of carrying out the method according to the invention, said liquid is selected and said method is performed such that the liquid will completely, or at least to 99%, leave the pellet before or during the sintering step. Since the liquid will leave the pellet, the liquid (and the material which constitutes the liquid) will not be present in the sintered pellet. Consequently, the liquid acts as a carrier of the additive and will not influence the properties of the produced pellet. Preferably, the liquid leaves the pellet during a step of heating the pellet. This can either be a separate heating step before the sintering step or the heating that is performed during the sintering step. The latter alternative has the advantage that no separate heating step is necessary. According to a further manner of carrying out the method according to the invention, said additive constitutes or includes a substance which causes said larger grains in the sintered pellet, wherein said substance is selected and the method is performed such that the substance completely, or at least to 90%, preferably to at least 95%, more preferred to at least 99%, leaves at least an outer portion of the pellet before and/or during the sintering step. Some additives may affect the neutron economy, i.e. they may absorb neutrons. On the other hand, as explained above, large grains, at least in the outer portion of the nuclear fuel pellet, are advantageous in order to prevent the HBS. It may therefore be an advantage to use a substance that leaves the pellet before or during the sintering step (but that causes the larger grains). This is also made easier by the present invention, since with the present invention, the additive may be added only to an outer portion of the pellet. If the additive is only present in an outer portion, it is easier to make the additive leave the pellet, for example during a heating step, such as the sintering step. It will now be explained why it is stated that the substance causes the larger grains. The substance may be included in a compound, such that the additive is a compound which includes the substance which causes the larger grains. The rest of the compound may act primarily as a carrier of the substance which causes the larger grains. For example, the additive may be UB4. In this case the B will cause the larger grains, but the U as such in the compound UB4 will not substantially contribute the larger grains. This is thus the reason why it is stated in the claim that the substance causes the larger grains. The substance is preferably a chemical element, for example B or Cr. The additive may include more than one such substance. According to a further manner of carrying out the method according to the invention, said additive is made of, or comprises, B and/or Cr. These materials are advantageous substances that will increase the grain size. The additive comprising B may for example be UB4, B4C, ZrB2 or just B. The additive comprising Cr may preferably be is the form of a chromium oxide, such as CrO, CrO2 and/or Cr2O3. According to a further manner of carrying out the method according to the invention, said additive comprises B and wherein at least 90% of said B is 11B. B in the form of the isotope 10B acts as a neutron absorber. However, if the purpose of the added B is to increase the grain size, but not to act as a neutron absorber, then it is preferable to use the isotope 11B, since if some B remains in the sintered pellet, this B will in this case not act as a neutron absorber. The B may, for example, be selected such that it in said additive to at least 98% is present in the form of the isotope 11B. According to a further manner of carrying out the method according to the invention, said liquid is selected such that the additive does not dissolve in the liquid, and such that the nuclear fuel material in the green pellet is not dissolved by the liquid. It is preferable that the liquid does not interact with the additive or the green pellet in such a manner that the additive or the green pellet is dissolved by the liquid. According to a further manner of carrying out the method according to the invention, said liquid is an oil, preferably a mineral oil. Such liquids have advantageous properties for acting as a carrier for the additive. Furthermore, by selecting a suitable mineral oil, a suitable viscosity is achieved. The invention also concerns a method of making and using nuclear fuel. This method comprises: making a plurality of nuclear fuel pellets according to any one of the preceding manners, arranging the nuclear fuel pellets in cladding tubes, arranging the cladding tubes, with the nuclear fuel pellets, in the core of a nuclear power reactor in a nuclear power plant, such that at least 20%, preferably at least 50%, most preferred 100%, of the nuclear fuel material in said core are made of pellets made in accordance with any one of the preceding manners, operating the nuclear reactor to produce energy. By using the advantageous nuclear fuel pellets obtained with the method according to the present invention in a real nuclear power reactor, the advantages of the produced nuclear fuel are thus achieved in a nuclear power reactor plant for producing energy. The nuclear power reactor preferably comprises several thousand cladding tubes comprising nuclear fuel pellets produced with the method according to the present invention. Since a person skilled in the art knows how to produce nuclear fuel pellets from a powder, all the details of such a method will not be described herein. However, the main steps which are relevant to the present invention are described. FIG. 1 shows schematically the main steps of a manner of carrying out the method according to the present invention. A nuclear fuel material in powder form is provided. The nuclear fuel material may be based on UO2, which is enriched concerning 235U. The powder may also comprise other materials, for example binder materials, U3O8, burnable neutron absorbers, pore formers and sintering aid. The powder may contain at least 60%, preferably at least 80%, UO2. The powder is pressed such that a “green” pellet is formed. The green pellet will be porous. For example 50% of the pressed pellet may consist of pores. An additive is provided. The additive is such that it will increase the grain size of the sintered pellet. The additive may for example comprise B (which will cause larger grains), for example in the form of UB4. According to one embodiment, the B is in the form of 11B. According to another alternative, the additive may be Cr2O3. The additive is preferably in the form of particles, i.e. a powder. The size of the particles should be small enough so that the particles can penetrate into the pores in the green pellet, into which it is intended that the particles should penetrate. The particle size may for example be about 1 μm. A liquid is provided. The liquid may be a mineral oil. The mineral oil may be selected to have a desired viscosity, for example a kinematic viscosity of 320 centistokes. The additive is mixed with the liquid. Preferably, the additive particles are dispersed in the liquid, i.e. the liquid is selected such that the additive particles do not dissolve in the liquid, and also such that the nuclear fuel material in the green pellet is not dissolved by the liquid. The green pellet is brought into contact with the liquid with the additive. The green pellet may for example be dipped into the liquid with additive or the liquid with additive may be sprayed onto the green pellet. The penetration depth of the liquid, and thereby of the additive, into the green pellet is controlled. This can be done by selecting a suitable viscosity of the liquid or by controlling the amount of liquid, with the additive, which is added to the green pellet. This can be done for example by spraying a certain amount of the liquid onto the pellet or by dipping the green pellet in the liquid, with additive, during a predetermined time. It is also possible to control the penetration depth by the particle size or the particle size distribution. According to one manner of carrying out the method according to the present invention, the penetration depth is controlled such that the additive is added only to an outer portion of the green pellet. By controlling for example the viscosity of the liquid, with the additive, or the size of the additive particles, it is also possible to control into which pores in the green pellet that the additive will enter. For example, it may be controlled that the additive will substantially only enter into the pores which exist between the grains in the green pellet. Alternatively, it may be controlled that the additive will enter also into the pores which exist in the grains in the green pellet. The so treated green pellet is then sintered. This can be done by a normal sintering process, for example in a furnace which contains different zones where the pellet is heated up to a final temperature of about 1800° C. The liquid is preferably selected such that it will evaporate during the heating process. There may be a separate heating step before the actual sintering in order to evaporate the liquid. However, no such separate heating step may be necessary, since the liquid will evaporate during the sintering process. According to one manner of carrying out the method according to the invention, also the substance that causes the larger grains in the sintered pellet will leave the pellet (evaporate) during a heating process, for example during the sintering step. When it is desired to increase the grain size in the whole nuclear fuel pellet, the additive, and the viscosity of the liquid, may be selected such that the whole pellet is infiltrated with the liquid with the additive. However, as explained above, it is possible to control the penetration depth of the liquid with the additive. According to a preferred manner of carrying out the present invention, the penetration depth is controlled such that the additive will substantially enter only into an outer peripheral portion of the green pellet. When the green pellet is then sintered, larger grains will be obtained mainly in an outer portion of the pellet. FIG. 2 illustrates schematically how the grain size may vary in a pellet treated in this manner. The x-axis shows the radius of the sintered pellet. The radius r 1.0 is thus the outer periphery of the pellet. The radius of the pellet may for example be about 4.6 mm. The y-axis in FIG. 2 shows the average 2D grain size. The curve in FIG. 2 thus shows how the average 2D grain size varies with the radius. FIG. 2 thus illustrates that according to this embodiment of the invention, a substantially larger grain size is obtained in the outer portion of the sintered pellet. This has in particular the advantage that the occurrence of the above described HBS can be prevented or delayed. A plurality of nuclear fuel pellets are produced according to the method of the present invention. The produced pellets are arranged in cladding tubes. The cladding tubes are arranged in the core of a nuclear power reactor, such that the core includes several thousand cladding tubes with pellets produced in accordance with the present invention. The nuclear reactor is operated in order to produce energy. The present invention is not limited to the examples described herein, but can be varied and modified within the scope of the following claims. |
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053496256 | abstract | An x-ray diagnostics installation for peripheral angiography examinations includes a control unit having an arithmetic unit that, on the basis of subject-related data supplied thereto, effects a pre-setting of the electrical parameters of the installation required for every exposure, plus the step length and the number of steps of the relative adjustment of the exposure unit and the patient support relative to one another, as well as the required diaphragm setting. By virtue of the pre-setting, stress on the patient is reduced, and since the overall time per examination is also reduced, patient throughput can be increased. |
046408134 | claims | 1. A soluble burnable absorber rod, comprising: (a) an elongated hollow tubular member having opposite ends and a hermetically sealed chamber defined therein between its said opposite ends, said tubular member including (b) a neutron absorber material in liquid form contained in said sealed chamber within said tubular member; (c) means providing a hydride sink disposed at one end of said tubular member and in communication with said sealed chamber, said sink being provided by one of said end plugs of said tubular member; and (d) means providing a hydrogen getter disposed at the other end of said tubular member and in communication with said sealed chamber, said means providing said getter being in the form of said liquid absorber material is boric acid enhanced with a B-10 isotope. (a) an elongated hollow tubular member having opposite ends and a hermetically sealed chamber defined therein between its ends; (b) said tubular member being formed by a tubular body of thin wall construction and a pair of end plugs attached to opposite ends of said body so as to hermetically seal the same and define said chamber in said member; (c) a quantity of neutron absorber material in liquid form partially filling said chamber, said material being in the form of boric acid enhanced with a predetermined concentration of B-10 boron isotope; (d) reinforcing convolutions formed in said body of said tubular member and extending between said ends thereof for strengthening said member against external pressure acting on said body; and (e) means on said tubular member and in communication with said sealed chamber therein providing a getter for hydrogen and a sink for hydride generated by the absorption of neutrons by said boric acid in said chamber, said hydrogen getter means being in the form of 2. The burnable absorber rod as recited in claim 1, wherein: 3. The burnable absorber rod as recited in claim 1, wherein said convolutions take the shape of a groove formed in said body so as to extend along a spiraling path between said ends of said tubular member. 4. The burnable absorber rod as recited in claim 1, wherein said convolutions take the shape of a series of circular grooves formed in said body so as to extend circumferentially about and be spaced axially along said tubular member between said ends thereof. 5. In a fuel assembly for a nuclear reactor including a plurality of guide thimbles and a plurality of nuclear fuel rods spaced apart from one another and from said guide thimbles and grouped together in an array organized to generate a neutron flux in said fuel assembly, an improved burnable absorber rod for insertion into at least one of said guide thimbles for regulating said reactor neutron flux, comprising: 6. The burnable absorber rod as recited in claim 5, wherein said hydride sink is in the form of one of said end plugs of said member. 7. The burnable absorber rod as recited in claim 5, wherein said convolutions take the shape of a groove formed in said body so as to extend along a spiraling path between said ends of said tubular member. 8. The burnable absorber rod as recited in claim 5, wherein said convolutions take the shape of a series of circular grooves formed in said body so as to extend circumferentially about and be spaced axially along said tubular member between said ends thereof. |
description | The present invention relates to a nondestructive inspection system using nuclear resonance fluorescence. Various isotopes including fissionable materials are contained in fuel rods and radioactive waste that are handled in nuclear fuel cycles for nuclear power generation. Inspecting these in a nondestructive manner and visualizing their spatial distribution are important in realizing safe and efficient nuclear fuel cycles. By way of example, with respect to cesium, there is Cs-133, which is a stable isotope with an atomic mass number of 133, and there is Cs-137, which is a radioactive nuclide with an atomic mass number of 137, where the handling of the latter is strictly controlled by law. In radioactive waste treatment, by quickly identifying them, costs associated with underground disposal can be reduced dramatically. The realization of a technique for identifying isotopes and visualizing the spatial distribution thereof is therefore strongly desired. In addition, in transporting nuclear fuel materials, measuring, in a nondestructive manner and from outside of a container, fissionable materials, such as uranium, etc., nuclear fuel materials, explosives, and materials from which they are made concealed inside the container is extremely important for purposes of strictly restricting/controlling transportation of nuclear materials, in preventing terrorism, etc., involving explosives, and in realizing a safe and secure society. Currently, at some nuclear fuel cycle facilities, seaports, airports, etc., X-ray transmission image-based internal geometry measurement and nondestructive inspection of fissionable materials, nuclear fuel materials, explosives, etc., are performed using large-scale X-ray inspection equipment that inspects fuel rods and containers in their entirety, or prompt gamma ray analyzers that use a neutron generator, etc. In X-ray inspection, a hard high-energy bremsstrahlung X-ray is used, and while there is an advantage in that a clear transmission image is obtained, the material cannot be identified. Prompt gamma ray analysis by way of neutron radiation allows for material identification and isotope identification, but has poor spatial resolution, and its spatial resolution is insufficient for interior visualization. The term material identification as used herein refers to element identification, that is, to the identification of atoms. This may be done by observing the electron state around the nucleus, and may be observed with relative ease by means of X-rays, etc. The term isotope identification refers to the identification of an isotope, that is, of a nucleus with a different number of neutrons, with respect to the protons and neutrons contained in the nucleus, and may be observed by detecting gamma rays. With respect to material inspection of the interiors of import container cargos and suit cases, there is proposed an isotope identification method based on nuclear resonance fluorescence (NRF) using bremsstrahlung X-rays (Patent Document 1). The nucleus of an isotope has, depending on the number of protons and neutrons, which are constituent elements thereof, a natural frequency (excitation level). When the isotope is irradiated with a photon having an energy that matches this frequency, the isotope absorbs the photon, and a fluorescent photon is thereafter generated upon deexcitation, and this is referred to as NRF. Isotope identification may be carried out by observing NRF gamma rays with a radiation detector. Since NRF gamma rays have an energy of several MeV and are capable of passing through an iron plate of approximately 10 mm, isotope identification for and the spatial distribution of a material sealed inside a container, etc., may be measured in a nondestructive manner. The concept of the method is shown in FIG. 1. A sample 2 is irradiated with photon beams 1 of X-rays, photons, etc. An isotope 3 of interest is contained in the sample 2. It is noted that the sample 2 may be shielded in some cases, however the shield is omitted here. The isotope 3 absorbs the photon beams 1, and emits an NRF gamma ray 4 which is detected by a radiation detector 6. The other photons are scattered by the other atoms within the sample to become scattered X-rays 5, and either exit the system or are detected by the radiation detector 6. A portion of the photon beams 1 that has been transmitted is measured at a photon intensity monitor 7. By scanning the photon beams 1 or by moving the sample 2, the spatial distribution of isotopes is measured. With respect to the above-mentioned inspection, analysis, or treatment process, there is proposed a method in which quasi-monochromatic photons are generated through laser-Compton scattering (LCS, later discussed) instead of bremsstrahlung X-rays, and in which these are used for isotope detection. It is possible to generate LCS photons in the photon range of several MeV by irradiating a high-energy electron beam, which is generated by an electron accelerator, etc., with high-intensity laser. As in ordinary Compton scattering, LCS is an interaction between electrons and photons, but is characterized in that the energy of the electrons is high, and that laser is used as photons. Photons generated by this method have the following characteristics: the photons are emitted within an extremely narrow solid angle and have high directionality comparable to synchrotron radiation; the photons may be made quasi-monochromatic by means of a collimator while at the same time reducing the energy spread (making them quasi-monochromatic) (Equation (1)) since there is a correlation between scattering angles of photons and energy; LCS photons with a high degree of polarization may be obtained since the polarization of the laser is preserved as is in the scattered photons (Equation (6), later discussed); and so forth. The principles of LCS are represented in FIG. 2, and the relationship between energy Eγ of LCS photons and energies of electrons and laser light through Equation (1). In Equation (1), Ee represents the energy of electrons, and EL the energy of laser light. E γ = E L ( 1 - β cos [ θ 1 ] ) 1 - β cos [ θ 2 ] + E L ( 1 - β cos [ θ 2 - θ 1 ] ) E e where β = 1 - γ 2 , γ = E e 0.511 ( 1 ) The relationship between energy and scattering angle of LCS photons with respect to cases where electrons with an energy of 641 MeV are irradiated with lasers whose wavelengths are 1064 nm and 1550 nm is shown in FIG. 3. By restricting scattering angle θ2, it is possible to obtain photons with the desired energy and energy width. Specifically, the scattering angle is restricted by positioning a collimator in which a narrow hole is opened in lead, etc., along the beam axis. The energy width is ordinarily on the order of several %, and these are referred to as quasi-monochromatic photons. An isotope's reaction cross-section σD(E) is given by Equation (2). Resonance width Γ is broadened due to Doppler broadening as given by Equation (3). However, the width of Δ is extremely narrow, and is ordinarily on the order several hundred meV. σ D ( E ) = π 3 2 ( ℏ c E ) 2 2 I 1 + 1 2 I 0 + 1 Γ Δ exp [ - ( E - E res Δ ) 2 ] ( 2 ) Δ = E res 2 kT eff mc 2 ( 3 ) Thus, in order to cause NRF efficiently, it is preferable that the excitation photon have a narrow energy spectrum that synchronizes with the natural frequency of the isotope. In the equations above, ℏ represents the Planck constant/2π, c the speed of light, E the photon energy, I0 and I1 the total angular momenta in the ground state and the excited state, respectively, Eres the resonance energy, Γ the resonance energy width, k the Boltzmann constant, Teff the effective temperature of a nucleus, and m the rest mass energy of an electron. Since the energy spread of LCS photons can be narrowed to or below several %, it is possible to increase the signal-to-noise ratio (S/N) by reducing background noise (noise), which is advantageous to the method using bremsstrahlung X-rays. Thus, measuring methods that employ an LCS photon beam are superior in many aspects, such as precision, time, reliability, safety, etc., over cases where a bremsstrahlung X-ray is used. In Non-Patent Document 1, there is proposed a method that uses LCS photons in an isotope identification method employing NRF. In Non-Patent Document 2, there is reported a method in which an Energy Recovery Linac (ERL), which is a next-generation electron accelerator, is combined with a state of the art high-power mode-locked fiber laser and a super cavity that accumulates pulsed laser, thereby generating LCS photons that are far more intense (approximately 108 times so) than existing LCS photons. It is indicated that the abundance of long-lived nuclides within radioactive waste can thus be detected in a few seconds. Non-Patent Documents 3 and 4 contain reports regarding the detection of an NRF gamma ray with an energy of 5512 keV generated from lead-208, which is an isotope of lead, and imaging based thereon, as well as the detection of a 4439 keV NRF gamma ray from carbon-12, and material identification based thereon. The lead-208 sample was carefully concealed within an iron box with a thickness of 1.5 cm. The nuclear excitation levels of carbon-12 and lead-208 are shown in FIG. 4. An LCS photon beam with a narrow energy width is used to excite nuclei. With the exception of hydrogen, there exists a unique excited state for each nucleus. When photons are emitted at the level (e.g., 5512 keV) of the nuclide to be measured (e.g., lead-208), the 5512 keV photons are absorbed by lead-208. In the process of cooling from the excited state, lead-208 emits NRF gamma rays that are equivalent to the excitation energy. By detecting these, it is possible to detect lead-208. LCS photon beams of the desired energy are generated by irradiating, with a laser with a wavelength of 1064 nm, 560 MeV electrons with respect to lead-208, and 510 MeV electrons with respect to carbon-12. As a method for accurately analyzing elements contained within a substance, there is X-ray fluorescence analysis. Since X-ray fluorescence is low in energy, substances concealed inside containers cannot be measured. Further, since element analysis is carried out by utilizing the fact X-rays generated due to the structures of atoms, that is, due to electron transition (characteristic X-rays), represent states unique to the atoms, while elements may be identified, isotopes may not. There is proposed a method in which a sample is irradiated with a high-energy gamma ray to induce a (γ, n) reaction with respect to the isotope of interest, thereby producing nuclear isomers, and in which isotope identification is carried out using the deexcitation gamma ray thereof (Patent Document 2). Patent Document 1: U.S. Pat. No. 7,286,638 Patent Document 2: JP 2004-219187 A Patent Document 3: JP 2006-318746 A Non-Patent Document 1: J. Pruet, D. P. McNabb, C. A. Hagmann, F. V. Hartemann and C. P. J. Barty, J. Appl. Phys. 99 (2006) 123102 Non-Patent Document 2: R. Hajima, T. Hayakawa, N. Kikuzawa, E. Minehara, J. Nucl. Sci. and Technol. Vol. 45, No. 5, pp. 441-451, 2008 Non-Patent Document 3: Nobuhiro Kikuzawa, Ryoichi Hajima, Takehito Hayakawa, Toshiyuki Shizuma, Eisuke Minehara, Hiroyuki Toyokawa, Hideaki Ohgaki: “Verifying the principles of a radioactive isotope detection technique by photonuclear resonant scattering using quasi-monochromatic γ-rays”, 2008 Annual Meeting of the Atomic Energy Society of Japan Non-Patent Document 4: Hideaki Ohgaki, Toshiteru Kii, Kai Masuda, Hiroyuki Toyokawa, Ryoichi Suzuki, Nobuhiro Kikuzawa, Toshiyuki Shizuma, Takehito Hayakawa, Ryoichi Hajima, Eisuke Minehara: “Material identification—light nucleus identification using nuclear resonant scattering by laser inverse Compton scattered γ-rays”, 2008 Annual Meeting of the Atomic Energy Society of Japan Imaging that utilizes prompt gamma rays by neutron radiation have the following problems in addition to its shortcoming of being poor in spatial resolution. Specifically, not only the substance within a container, but also the nuclei of the iron and lead with which the container is built are excited at the same time due to neutron radiation. The prompt gamma ray spectra consequently become extremely complex, which makes solving them within a short period extremely difficult. In addition, when neutrons are absorbed by the nuclei of uranium-233 or -235, plutonium-239, uranium-238, etc., nuclear fission could occur. In particular, uranium-233 and -235, and plutonium-239 pose a high risk of triggering a chain reaction. In addition, since the atomic number sometimes changes due to neutron radiation, there is also a problem in that it may turn into a substance that is different from the original substance (nuclear transmutation). Since bremsstrahlung X-rays vary widely in wavelength, numerous levels of a plurality of isotope nuclei may be excited simultaneously. However, the energy width that contributes to NRF is extremely narrow and on the order of several hundred meV even with thermal Doppler broadening taken into account, and all X-rays with energies falling outside of that become noise, thereby giving rise to a problem which is that S/N becomes extremely small. Specifically, the fact that reliability with respect to measurement precision, quantifying capability, etc., is low, and the fact that measurement takes a long time are practical problems. In addition, when the electron energy is made to be equal to or higher than the neutron emission energy (generally around 8 MeV, with some variation depending on the nuclide), neutrons are emitted from the isotope to be measured, which causes numerous problems, such as safety, increased costs for installing shielding, and so forth. In particular, if a fissionable material is contained in the sample, there is a risk of nuclear fission being triggered. Therefore, using bremsstrahlung X-rays equal to or above the neutron emission energy is extremely dangerous and must be avoided at all costs. As the electron energy is lowered, the energy spectrum of bremsstrahlung X-rays as a whole shifts towards the low-energy side, and the neutron emission rate drops. In the low-energy region, large amounts of low-energy X-rays and characteristic X-rays are emitted due to atomic scattering. There is consequently a disadvantage in that NRF gamma rays to be measured become buried in noise and S/N becomes small. In particular, it has been difficult to detect important isotopes such as carbon-12, nitrogen-14, etc., whose excitation levels are at 4 to 5 MeV. By collimating an LCS photon beam to a diameter of several mm, it is possible to narrow the energy width to or below several %. In Patent Document 1, Non-Patent Document 1 and Non-Patent Document 2 discussed above, it is mentioned that S/N is improved utilizing this method, while also shortening measurement time. However, with the methods presented in these documents, since only a narrow energy width can be excited, it is not possible to inspect a plurality of elemental ratios in one measurement. In view of the above, a method of inferring the type of an explosive by measuring the abundance ratios and abundances of isotopes nondestructively and remotely is proposed. By way of example, in naturally occurring carbon, 98.89% of carbon-12 (12C) is contained, and 99.64% of nitrogen-14 (14N) and 99.76% of oxygen-16 (16O) are respectively contained in nitrogen and oxygen. As levels of these isotopes, carbon-12 (4439 keV), nitrogen-14 (4915 keV and 7029 keV), and oxygen-16 (6917 keV and 7117 keV) are used. As the levels above are spaced apart in energy, only a single level can be excited with conventional methods. Thus, there was a disadvantage in that, even if the abundance of carbon-12 were detected with high precision, it was impossible to tell whether it was from an explosive or from surrounding structures. In other words, although there is a need to vary the LCS photon energy to measure different isotopes, because conventional methods only allow for excitation of a single level, the electron energy or laser wavelength had to be varied. With respect to LCS photons, since a laser and an electron beam are made to collide after being converged to a size of several tens of μm to 100 μm in diameter, yield would drop significantly if the electron orbit were to be off even slightly. Although it is not impossible to adjust the orbit of an electron beam with precision on the order of several tens of μm using existing techniques, it would require extensive measurement apparatus and extended periods of adjustment, thereby restricting implementation thereof to large-scale accelerators. When the energy of an electron beam is varied, the electron orbit varies, and it becomes difficult to generate LCS photon beams along the same axis, thereby precluding accurate visualization of the positions of isotopes. The fact that extended periods are required for, in addition to such fine adjustments, fine adjustments of the position of the collimator positioned along the laser or beam axis is also a practical problem. A nondestructive inspection system of the present invention comprises: a plate that holds a sample; a plurality of radiation detectors positioned in predetermined directions relative to the sample held by the plate; a photon beam irradiation part that irradiates the sample held by the plate with a plurality of types of quasi-monochromatic photon beams having differing energies and whose planes of polarization are each controlled, the quasi-monochromatic photon beams being made to be coaxial; a drive part that relatively moves the plate and the photon beam irradiation part; a control processing part that controls the drive part and to which detection signals from the plurality of radiation detectors are inputted; and a display part, wherein the radiation detectors are positioned in directions that allow for detection of NRF gamma rays emitted from a nuclear level of an isotope within the sample in directions that are dependent on the planes of polarization of the quasi-monochromatic photon beams with which the sample is irradiated, wherein the control processing part identifies an isotope of interest that is present in a photon beam irradiated region in the sample based on the detection signals of the radiation detectors, and wherein a spatial distribution thereof is visualized and displayed on the display part. LCS photon beams are suitable for use as the plurality of types of quasi-monochromatic photon beams whose planes of polarization are each controlled. However, other photon beams having similar polarization characteristics and energy characteristics may also be used. By way of example, the photon beam irradiation part utilizes laser Compton scattering to coaxially generate a plurality of LCS photon beams having a single energy or several energies and whose planes of polarization are each controlled. Specifically, the photon beam irradiation part comprises: an electron beam accelerator that generates an electron beam of a predetermined energy; a first laser light source that generates laser light of a first wavelength and having a first polarization; and a second laser light source that generates laser light of a second wavelength and having a second polarization, wherein the laser light of the first wavelength and the laser light of the second wavelength are made to collide with the electron beam at a first angle and as coaxial light. A first polarized LCS photon beam resulting from the laser light of the first wavelength and a second polarized LCS photon beam resulting from the laser light of the second wavelength, which are generated by laser Compton scattering, are then extracted via a collimator positioned in a second angular direction relative to the electron beam. It is preferable that the first laser light source and the second laser light source be controlled by the control processing part and that they each generate pulsed light with a different time structure. The planes of polarization of the LCS photon beams may be made to be mutually parallel or perpendicular to the plane of polarization of one LCS photon beam thereamong. In addition, by way of example, the radiation detectors are positioned, with respect to the sample, in a direction parallel to and/or a direction perpendicular to the planes of polarization of the LCS photon beams with which the sample is irradiated. When exciting a certain nuclear level of the isotope of interest within a sample, the levels of other nuclides and isotopes sometimes exist in close proximity thereto. In such cases, as a plurality of NRF gamma rays are generated simultaneously, they may sometimes be indistinguishable in an energy spectrum. As a method of dealing with such cases, a method has been invented wherein the emission directions of NRF gamma rays are controlled by switching the planes of polarization of LCS photon beams, and wherein the NRF gamma ray of interest is detected with high precision by emitting each of the NRF gamma rays in an anticipated direction. Specifically, the LCS photon beams with which the sample is irradiated are polarized, and the radiation detectors are positioned in certain angular directions relative to their planes of polarization. Although NRF gamma rays are emitted from the sample, each is emitted in a certain direction relative to the plane of polarization of the LCS photons in accordance with the type of transition. Thus, an energy spectrum may be measured with high efficiency and high precision using a plurality of radiation detectors with respect to NRF gamma rays emitted from a plurality of isotope and nuclear levels whose energy levels are in close proximity to one another. In addition, since radiation detectors with high energy resolution ordinarily have a slow response, the upper limit for count rates applicable to measurement is low. However, by using the present method, highly precise measurements are possible even when detectors with low energy resolution are used, and it is thus possible to accommodate high count rate measurements. In other words, by dispersing the emission directions of the NRF gamma rays emitted from the nuclear levels of the isotopes within the sample in a dependent manner with respect to the planes of polarization of the photon beams with which the sample is irradiated, it is possible to lower the count rate per radiation detector. With the present invention, isotope identification imaging of nuclear fuel materials or explosives concealed within drums or containers in which nuclear reactor fuel or radioactive waste is sealed can be realized while ensuring high precision, high reliability and safety. Embodiments of the present invention are described below with reference to the drawings. FIG. 5 is a schematic view showing a system configuration example for practicing a nondestructive inspection method according to the present invention. This system identifies isotopes within a sample by NRF using LCS photons, and images and displays the spatial distribution thereof. A nondestructive inspection device of the present embodiment comprises, as an LCS photon beam irradiation part, an electron beam accelerator 11, laser light sources 13 and 17, a collimator 23, radiation detectors 32 and 33 that detect NRF gamma rays generated from a sample, and a photon intensity monitor 34 for monitoring the intensity of photon beams that have penetrated the sample. A sample 31 is positioned on a movable plate 35. By moving the movable plate 35 two-dimensionally in directions parallel to the plane of the paper and in directions perpendicular to the plane of the paper by means of a drive part 36, it is made possible to scan the sample 31 two-dimensionally with respect to the LCS photon beams with which it is irradiated. The laser light sources 13 and 17, and the drive part 36 are controlled by a control processing part 37. The outputs of the radiation detectors 32 and 33, and of the photon intensity monitor 34 are processed at the control processing part 37 to detect, with respect to each move position of the movable plate 35, the abundance of a single or a plurality of isotopes of interest in the sample 31, while at the same time quantifying the abundance ratio thereof to image a spatial distribution and display it on the screen of a monitor 38. The radiation detectors 32 and 33, and the photon intensity monitor 34 may be placed on the movable plate 35. FIG. 5 shows such a case. With respect to an electron beam 12 of a predetermined energy generated from the electron beam accelerator 11, laser light 16 and 20 of the same wavelength or of a plurality of types of distinct wavelengths (hereinafter referred to as multiple-wavelength) are generated from the laser light sources 13 and 17. The laser light 16 and 20 have their planes of polarization adjusted by wave plates 14 and 18. By way of example, it is assumed that the laser light 16 is vertically polarized with its electric field vector lying in a direction perpendicular to the plane of the paper, and that the laser light 20 is horizontally polarized with a plane of polarization that lies in a plane that is parallel to the plane of the paper. Instead of being linearly polarized, the plane of polarization of the laser light may be circularly polarized or randomly polarized as well. The polarized laser light 16 and 20 are deflected by mirrors 15 and 19, and are made to collide with the electron beam 12 at angle θ1 as shown in FIG. 2. For the mirror 15, by using a dielectric multilayer mirror, etc., with a narrow wavelength band, it can be made transparent with respect to the laser light 20, thereby making it possible to make the two laser light 16 and 20 coaxial. Although not described herein, using a focusing optical element, etc., each laser light is focused at a point at which it is made to collide with the electron beam 12, thereby effectively generating an LCS photon beam. By having the laser light 16 and 20 irradiate the electron beam 12 in a pulsed manner with respectively different time structures, such as pulse frequencies or generation timing, etc., pulsed LCS photon beams 21 and 22 are generated at the same scattering angle of θ2. The pulsed LCS photon beams 21 and 22 may differ from each other in energy. As will be discussed later, in LCS, the plane of polarization of a laser becomes, as is, the plane of polarization of LCS photons (Equation (6), discussed below). It is thus possible to generate a high-energy photon beam whose polarization may be controlled as desired. The pulsed LCS photon beams 21 and 22 with differing planes of polarization are, after having their energies and spatial distributions shaped by the collimator 23, made to irradiate the sample 31, thereby generating NRF gamma rays 39 and 40. It is assumed that only the polarized pulsed LCS photon beam 21 is capable of exciting a reaction that generates the NRF gamma ray 39 of interest of an isotope of interest contained in the sample 31. It is similarly assumed that only the polarized pulsed LCS photon beam 22 is capable of exciting a reaction that generates the NRF gamma ray 40 of interest of another isotope of interest contained in the sample 31. The isotopes may be of different types or of the same type, but the types of transition of the excitation levels must be different. Types of transition are discussed later. Due to factors stemming from the nuclear structure, the NRF gamma ray 39 is emitted only in the direction of the radiation detector 32, and is detected by the radiation detector 32. In this case, no signal is generated at the radiation detector 33. The NRF gamma ray 40 is emitted only in the direction of the radiation detector 33, and is detected by the radiation detector 33. In this case, no signal is generated at the radiation detector 32. It is noted that the photon intensity monitor 34 should preferably be located at a part of the beam furthest downstream in order to monitor the intensity of the emitted LCS photon beam. By using the photon intensity monitor 34, it is possible to find the absolute values of the intensities of the NRF gamma rays 39 and 40. It is thus possible to quantify the isotope abundance ratio within the sample. It is assumed here that the interaction cross-section and energy of NRF are known. By simultaneously moving the pulsed LCS photon beams 21 and 22 and the collimator 23 to scan over the sample 31, or by moving the sample 31, it is possible to image the spatial distribution of a plurality of isotopes of interest. In the present embodiment, by driving the movable plate 35, on which the sample 31 is placed, by means of the drive part 36 and relative to the collimator 23, the sample 31 is scanned with the pulsed LCS photon beams 21 and 22. In addition, by appropriately adjusting the power of the laser light sources 13 and 17, it is possible to adjust the intensity of the pulsed LCS photon beams 21 and 22, and to quantify the abundance ratios of a plurality of isotopes contained in the sample 31. With the present invention, by irradiating the same electron beam with laser light of a single or a plurality of wavelengths, single-wavelength or multiple-wavelength LCS photon beams with aligned beam axes and divergence angles are generated, and a single or a plurality of levels of a single or a plurality of isotopes is/are selectively and simultaneously excited therewith. In JP 2006-318746 A, there is discussed a method of generating quasi-monochromatic X-rays of 100 keV or below at multiple wavelengths by LCS using a small electron accelerator and pulsed lasers with a plurality of types of distinct wavelengths for use in medical imaging, such as cardiac catheterization, etc. However, an object of the present invention is to generate high energy photon beams in order to quantify, with high precision and in a safe manner, the abundances, spatial distribution, and abundance ratio of isotopes concealed in a strictly shielded box whose maximum iron thickness may measure up to several cm, such as a container, a fuel cladding tube, etc., and to detect them with high efficiency through temporal/spatial discrimination per pulse utilizing the anisotropy of nuclear reactions. It thus differs entirely in concept and methodology. In addition, it is important that no neutrons be generated due to LCS photon beam irradiation, or that the generated amount thereof be suppressed to a tolerable level. Depending on the isotope, there is a risk that nuclear fission may be triggered due to photonuclear reactions, or a possibility that an isotope within the sample may become a different element through nuclear transmutation. There is also an advantage in that the cost at the time of construction can be reduced significantly by simplifying the shielding required for the measuring system as a whole. Thus, it is preferable that the LCS photon energy be made lower than the neutron emission energy, and that NRF of or below the neutron emission energy be utilized. With the present invention, by controlling the planes polarization of the emitted lasers, detection sensitivity for a certain level of an isotope of interest is improved. Some nuclear levels are such that NRF gamma rays are emitted in certain directions based on the plane of polarization of the photon beam with which excitation is carried out. By way of example, in electric dipole (E1) transition, magnetic dipole (M1) transition, electric quadrupole (E2) transition, etc., NRF gamma rays are emitted only in directions that are perpendicular or parallel to the electric field or magnetic field, respectively. Thus, by positioning a plurality of photon detectors, each having an appropriate energy resolution and directionality, above, below, to the left of, to the right of, etc., the unit under inspection, and by adjusting the planes of polarization of the excitation photon beams in advance in accordance with the desired level, it is possible to detect the desired NRF gamma rays with sufficient precision and spatial discrimination. A characteristic of Compton scattering is that the polarization of the photons before and after scattering is preserved. This will be explained by describing the degree of polarization of electromagnetic waves using the Stokes parameters (ax, ay, az). The direction in which laser light travels is taken to be the z-axis, and the x- and y-axes are so defined as to be perpendicular thereto. The degree of linear polarization of the laser light is denoted P1i, and is given by Equation (4).P1i=ax2−ay2 (4) In the electron rest frame, laser light that has undergone Compton scattering at scattering angle θ, that is, LCS photons, scatters at arbitrary azimuths relative to the rotation about the z-axis. Thus, degree of polarization P1f of the LCS photons becomes an average value integrated in all azimuths. This is represented in Equation (5). P 1 f = P 1 i ( 1 - cos θ ) 2 2 [ ( K 0 - K ) 2 K 0 K 1 + cos 2 θ ] ( 5 ) Here, K0 and K represent the energies of the laser and LCS photons in the electron rest frame. When expressed in the laboratory frame, Equation (5) would resemble Equation (6). P 1 f = P 1 i ( 1 - 1 - ρ ( 1 + a ) 1 - ρ ( 1 - α ) ) 2 2 [ ρ 2 ( 1 - a ) 2 1 - ρ ( 1 - a ) + 1 + ( 1 - ρ ( 1 + a ) 1 - ρ ( 1 - a ) ) 2 ] ( 6 ) Here, it is assumed that Kf and Kf—max are the energy and the maximum energy of the LCS photons in the laboratory frame, and that ρ is the ratio thereof as follows: ρ = K f K f _ max ( 7 ) a represents the degree of energy multiplication of the laser light with respect to the maximum energy of the LCS photons, and is given by Equation (8). a = K f _ max 4 γ 2 K i ( 8 ) The degree of polarization of the LCS photons becomes extremely close to 1 when θ is close to 180 degrees, that is, in the vicinity of ρ=1. Thus, it can be seen that, in LCS, the polarization of the laser light is preserved in the LCS photons, and that a degree of polarization of 100% can be attained near the center of the beam axis. In “Observation of M1 Resonance in 206Pb Using a Highly Linear Polarized Photon Beam”, NUCLEAR PHYSICS A649, pp. 73c-76c, 1999, this method is used to verify the fact that the level of lead-206 with unknown parity is M1. An object of the present invention is to detect isotopes and measure abundance ratios accurately by focusing on sufficiently reliable isotope levels listed in nuclear data tables, etc., and selectively exciting them. The present invention is of practical importance in that, even in cases where an unknown element is present in a sample of interest and where the level of that element is in proximity to the isotope of interest, so long as the parities, quantum numbers, etc., are different, they can be discriminated with sufficient precision. Specifically, by positioning detectors in directions in which NRF gamma rays are emitted, and by controlling the planes of polarization of the LCS photons, it is possible to, with high reliability, identify isotopes, quantify abundance ratios, and perform imaging thereof even if an isotope whose level is in proximity is present. The present invention differs from conventional methods in that an object thereof is to improve detection accuracy dramatically. Thus, in transporting nuclear fuel materials, it becomes possible to detect, from outside of a container and in a nondestructive, remote, and safe manner, fissionable materials or nuclear fuel materials, such as uranium, etc., explosives and substances from which they are made, or any other substance besides hydrogen concealed inside the container, and to measure the spatial distribution of isotope nuclei with high spatial resolution. Such objects as strictly restricting and managing the transport of nuclear materials, and a safe and secure society can thus be achieved by preventing terrorism, etc., involving explosives. In addition, it becomes possible to nondestructively and remotely inspect fissionable materials and various isotopes contained in fuel rods and radioactive waste that are handled in nuclear fuel cycles for nuclear power generation, and to visualize the spatial distribution thereof with high precision, thereby making it possible to attain a safe and efficient nuclear fuel cycle. Looking at the chemical formulae for typical industrial and military-use explosive substances used in the world today, it can be seen that most explosives contain carbon, oxygen and nitrogen, and that the proportion of carbon, oxygen and nitrogen varies depending on the explosive. The excitation levels of elements (12C, 14N, 16O) used in the detection of explosives are shown in FIG. 6. In the present example, carbon-12, oxygen-16 and nitrogen-14 are detected using two-wavelength LCS photon beams, and the abundance ratio thereof is measured. The excitation levels 6917 keV and 7117 keV of oxygen-16, 4915 keV and 7029 keV of nitrogen-14, and 4439 keV of carbon-12 are looked at. Here, by irradiating electrons of an energy of 641 MeV with lasers of wavelengths 1064 nm and 1550 nm, LCS photons of two different energies, namely 7250 keV and 5000 keV (energy spread approximately 12%), are generated, thereby exciting all of these levels. An energy spectrum of two-wavelength LCS photon beams generated by such a method is shown in FIG. 7. Since LCS photon beams are generated in the travel direction of the electron beam at divergence angles determined by the electron beam, so long as they share the same electron beam, LCS photon beams of differing wavelengths would also be generated along the same axis, and their properties as beams would generally be the same. For the laser with a 1064 nm wavelength, a Nd:YAG or Nd:YVO4 laser is used, and for the laser with a 1550 nm wavelength, a pulsed fiber laser doped with Er is used. It is thus possible to generate the desired two-wavelength LCS photon beams along the same beam axis, and to detect the NRF gamma rays of the respective isotopes. It is noted that, for the lasers, it is also possible to use an appropriately mixed second harmonic, third harmonic, or a higher harmonic or fundamental wave. The lasers are all perpendicularly polarized. As shown in FIG. 8, the radiation detectors are positioned above, below, to the left and to the right, that is, in directions perpendicular to and in directions parallel to the planes of polarization of the laser light used to generate LCS photon beams 51 and 52. If the laser light used to generate the LCS photons are linearly polarized, the LCS photon beams 51 and 52 would also be linearly polarized. The NRF gamma rays 53 of 4439 keV of carbon-12, of 6917 keV of oxygen-16, and of 7029 keV of nitrogen-14 are detected by the radiation detectors 64 above and below as shown in FIG. 8(a), while the NRF gamma rays 54 of 4915 keV of nitrogen-14 and of 7117 keV of oxygen-16 are detected by the radiation detectors 66 to the left and right as shown in FIG. 8(b). The respective NRF gamma rays 53 and 54 are generated with different laser pulses, and have temporally distinct structures. With respect to the radiation detectors, an energy resolution of 100 keV or below would be sufficient. The energy resolution of high-purity germanium detectors is approximately 10 keV at 5000 keV, which is sufficient for the present measurements. The intensity of the emitted LCS photon beam is measured by a photon intensity monitor 65 positioned behind a sample 61, but in order to improve measurement precision, it is synchronized with the laser pulse to reduce the influence of background radiation, etc. In the case above, it is, naturally, possible to omit either the radiation detectors 64 or 66 by switching the planes of polarization of the LCS photon beams to a perpendicular or horizontal direction. The isotope abundance ratios of carbon-12, oxygen-16 and nitrogen-14 are 98.89%, 99.762% and 99.634%, respectively. Thus, their respective abundance ratios may be measured based substantially on NRF gamma ray intensity ratios. Various explosives, carbon-to-nitrogen ratios (C/N), and oxygen-to-nitrogen ratios (O/N) are shown in FIG. 9. Through application of the present invention, it is possible to identify an explosive's type. It is noted that, through application of the present invention, it is possible to detect not only explosives, but also any kind of substance concealed in a container, etc., such as radioactive waste, nuclear fuel, banned drugs or substances from which they are made, etc. FIG. 10 is a diagram showing a concealed object being detected according to the present invention. It is now assumed that isotopes of a given nucleus are inside a thick shield box 71 made of iron as a concealed object 72. LCS photon beams are emitted theretowards. LCS photon beam A penetrates the shield box 71 without causing any interaction. LCS photon beam B causes NRF at the concealed object (isotopes of a nucleus) within the shield box, and an NRF gamma ray 74 thereof undergoes height discrimination at a detector 73. Another LCS photon beam, C, causes NRF at the shield box 71, and an NRF gamma ray 75 thereof also undergoes height discrimination at the detector 73. Since the NRF gamma ray 75 is NRF from a constituent element of the shield box, it has an energy that differs from that of the NRF gamma ray 74. Thus, the two are readily discernible using a pulse-height discrimination method. Using an excitation level that is at 4439 keV of carbon-12, which is a type of carbon isotope, experiments to verify the present method were attempted. A carbon isotope (carbon-12) is concealed in an iron box that is 1.5 cm thick. A horizontally polarized LCS photon beam with an energy of 4.6 MeV is emitted theretowards from the outside. The energy of the electron beam and the wavelength (polarization direction) of the laser light in this case are 510 MeV and 1064 nm (horizontal), respectively. The scattering directions of the NRF gamma rays were cut out with an angle spread of approximately 330 microradians about the 180-degree direction. The energy spread resulting therefrom is approximately 4.3% full width at half maximum. Since it is known that the 4439 keV level of carbon-12 undergoes an E2 transition, NRF gamma rays are emitted in directions parallel to the planes of polarization of the LCS photon beams. FIG. 11 is an energy spectrum of NRF gamma rays of 4439 keV of carbon-12 as measured by a high-purity germanium detector positioned in the horizontal direction. As can be seen therein, a clear peak was observed at 4439 keV. Based on the experiment above, results of isotope imaging measurements using NRF gamma rays are shown in FIG. 12. The shield box shown in FIG. 10 containing the concealed object was placed on the movable plate of the nondestructive inspection system described in connection with FIG. 5, and the shield box was linearly scanned with LCS photon beams. A heavy element was selected for the concealed object to simulate a nuclear material, Pb-208, which is an isotope of lead, was used, and 5512 keV was used as the excitation level. It is known that this level undergoes an E1 transition, and NRF gamma rays are emitted in directions perpendicular to the planes of polarization of the LCS photons. Radiation detectors were therefore placed in the horizontal direction, while the planes of polarization of the LCS photon beams were made to be in the perpendicular direction. Pb-208 was concealed in an iron box with a thickness of 1.5 cm, and perpendicularly polarized LCS photon beams with an energy of 5.7 MeV were emitted. The energy of the electron beam, and the wavelength and plane of polarization of the laser light in this case were 570 MeV, 1064 nm, and in the perpendicular direction, respectively. The scattering directions of the NRF gamma rays were cut out with an angle spread of approximately 330 microradians about the 180-degree direction. The energy spread resulting therefrom is approximately 7% full width at half maximum. High-purity germanium detectors were used for the radiation detectors, and after the energy spectrum was measured for approximately 30 minutes, the sample on the movable plate was moved in the up/down direction. At the position to which it was moved, an energy spectrum measurement was performed again, and by repeating the above, the correlation was measured between NRF gamma ray count of 5512 keV and sample position. The diameter of the LCS photon beams is 3 mm at the sample position. As can be seen from FIG. 12, NRF gamma rays of the lead isotope Pb-208 were strongly detected within the range of ±10 mm from the reference position. The spatial distribution of the concealed object was thus successfully imaged one dimensionally using the NRF gamma rays of the lead isotope Pb-208 concealed in an iron box with a thickness of 15 mm. High-purity germanium detectors cannot be used at high count rates, and the maximum count rate is approximately 10 kHz or below. In cases of high count rates, that is, where the NRF gamma ray intensity is high, measurements cannot be made with high precision. Shortening the duration of inspection by employing high count rate gamma ray detectors is important. By way of example, a case is considered where a screening inspection for dangerous objects with respect to 10 m long shipping containers is to be performed on all containers unloaded at a port. According to Roger Bostelman, “Requirements: Cargo Container Transfer Requirements for the Mobile Offshore Base”, Intelligent Systems Division, National Institute of Standards and Technology, 1998, assuming the per-hour throughput of a container crane is approximately 30 containers/hour, the container movement speed during inspection would be approximately 10 cm/s. Further, according to Brian Lewis, “PORT SECURITY: CONTAINER INSPECTION TECHNOLOGY, 2002 Technical Report”, The Logistics Institute, Georgia Institute of Technology, 2002, it is stated that the maximum inspection speed with respect to container-transporting trucks for ground transportation is 4-5 miles/h. Accordingly, an inspection speed of approximately several tens of cm/s, and up to 200 cm/s, would be required. If screening in which LCS photon beams with a diameter of 1 mm are emitted every 1 cm were to be performed, the irradiation time per point would be 0.005 to 0.1 seconds. While a detailed study is described in Non-Patent Document 2, by using LCS photon beams of an intensity that is currently technically feasible to a sufficient extent, isotope identification can be performed in 0.1 seconds or less, and it is thus apparent that the present invention is applicable to the inspection of freight containers. Since the maximum count rate anticipated in such cases is 10 kHz or greater, high-purity germanium detectors cannot be used. However, the present invention allows for the use of a scintillation detector that uses the inorganic scintillation material LaBr3(Ce), which accommodates count rates that are up to approximately three figures greater than high-purity germanium detectors (10 MHz). The energy resolution is 1 to 3%, which is 1/10 or less of that of high-purity germanium detectors, but is sufficient for identifying explosives. Since the planes of polarization of the LCS photon beams preserve the planes of polarization of the incident lasers as is, it is possible to control the directions in which NRF gamma rays are emitted. Consequently, the measurement precision, inspection throughput, and cost performance are improved. Specifically, by scattering NRF gamma rays with close energy levels in different directions in accordance with the difference in the planes of polarization, it is possible to construct an inspection system with which sufficiently high precision and inspection throughput may be attained even when cheap detectors with low energy resolution, such as scintillators, etc., are used. Nondestructive inspection according to the present invention may, in practice, be performed by being divided into first-stage screening and second-stage inspection. The former refers to measuring the abundance ratios of key elements (e.g., carbon/nitrogen, oxygen/nitrogen, uranium/thorium, etc.) by irradiating one spot or several spots with LCS photon beams, and the likelihood of it being a dangerous explosive or nuclear material is determined in a short period of time. In the latter, imaging inspection with more detailed isotope identification is performed through beam scanning. Next, an example in which first-stage screening and second-stage inspection are performed is described. In the present example, too, LCS photon beams of several energies are generated along the same beam axis, and the isotope abundance ratios of such elements as carbon, nitrogen, oxygen, etc., are measured by NRF. Further, precision was improved with respect to the discrimination method for the NRF gamma rays thereof. As for the device configuration of the nondestructive inspection system, that shown in FIG. 5 was used. For the radiation detectors 32 and 33, scintillation detectors having a count rate of 10 kHz or above were used although their energy resolution of 1 to 3% is inferior to high-purity germanium detectors. The device shown in FIG. 5 may be used for both the first-stage screening and the second-stage inspection. However, it is assumed that, in the first-stage screening, the sample 31 is irradiated with LCS photon beams in a relatively coarse manner, such as at predetermined intervals, etc., and that two-dimensionally scanning the LCS photon beams with which the sample 31 is irradiated is only performed in the second-stage inspection. In the second-stage inspection, the outputs of the radiation detectors 32 and 33, and of the photon beam intensity monitor 34 are processed at the control processing part 37, and with respect to each move position of the movable plate 35, the abundances of the plurality of isotope nuclei of interest within the sample 31 are detected, while at the same time the abundance ratios thereof are quantified to image the spatial distribution and display it on the screen of the monitor 38. With respect to the electron beam 12 of a predetermined energy generated from the electron beam accelerator 11, laser light 16 and 20 are generated from the laser light sources 13 and 17. The planes of polarization of the laser light 16 and 20 are adjusted by the wave plates 14 and 18. This may be attained readily by inserting, taking out or rotating a ½ wave plate. The planes of polarization are determined in accordance with the constituent elements of the object to be inspected or with the type of isotope to be inspected. The polarized laser light 16 and 20 are respectively deflected by the mirrors 15 and 19, and are made to collide with the electron beam 12 at angle θ1 as in FIG. 2. Here, in order to maximize photon energy, θ1 was defined to be 180 degrees. As discussed above, by using a dielectric multilayer mirror, etc., with a narrow wavelength band for the mirror 15, it is possible to coaxially irradiate the electron beam with the two laser light 16 and 20. As a result, it is possible to, without altering the properties of the electron beam at all, generate LCS photon beams with two types of energies by irradiating the same electron beam with lasers of two types of wavelengths. The intensity of each laser light is adjusted as deemed appropriate depending on the reaction cross-section (likelihood of interaction) of the isotope to be measured. It is preferable that it be so adjusted that the intensity of the emitted NRF gamma ray would generally have a variation of within approximately several tens of %. The laser light 16 and 20 generate the pulsed LCS photon beams 21 and 22 by irradiating the electron beam 12 in a pulsed manner with respectively different time structures, such as pulse frequencies or generation timing, etc. The pulsed LCS photon beams 21 and 22 may respectively differ in energy, or they may differ in terms of their planes of polarization, or they may differ in both. The pulsed LCS photon beams 21 and 22 with differing planes of polarization pass through the collimator 23, after which they irradiate the sample 31, and the NRF gamma rays 39 and 40 are generated. When the position and direction of the electron beam do not change, there is no need to adjust the collimator 23. In the present example, since the adjustment of the LCS photon beam energy is performed solely by way of the lasers, position adjustment of the collimator need only be performed once before starting measurement, and no subsequent adjustment is required at all. By utilizing LCS, it is possible to generate LCS photon beams with an energy spread of approximately 0.1 to 20% at high intensities. It is assumed that only the polarized pulsed LCS photon beam 21 is capable of exciting a reaction that generates the NRF gamma ray 39 of interest of an isotope of interest contained in the sample 31. It is similarly assumed that only the polarized LCS photon pulsed beam 22 is capable of exciting a reaction that generates the NRF gamma ray 40 of interest of another isotope of interest contained in the sample 31. The isotopes may be of different types or of the same type, but the types of transition of the excitation levels must be different. The NRF gamma ray 39 is emitted only in the direction of the radiation detector 32, and is detected by the radiation detector 32. In this case, no signal is generated at the radiation detector 33. The NRF gamma ray 40 is emitted only in the direction of the radiation detector 33, and is detected by the radiation detector 33. In this case, no signal is generated at the radiation detector 32. In addition, by using the photon intensity monitor 34 located at a part of the beam furthest downstream, it is possible to find the absolute values of the intensities of the NRF gamma rays 39 and 40. It is thus possible to quantify the isotope abundance ratio within the sample. The NRF gamma ray energies (keV) and transition models of carbon-12, nitrogen-14, and oxygen-16 used for the detection of explosives are shown in Table 1. TABLE 1ElementEnergyTransition modelCarbon-124439E2Nitrogen-147029M14915E1Oxygen-166917E27117E1 In Table 1, the symbols E1, M1, E2, etc., represent transition models of nuclei, where E1 signifies that NRF gamma rays are emitted in a direction perpendicular to the plane of polarization of the photon beam, and E2 and M1 signify that NRF gamma rays are emitted in a parallel direction. The excitation levels of the elements provided above are broadly in the ranges of 4.4 to 5 MeV, and 6.9 to 7.1 MeV. Thus, by generating two-wavelength LCS photon beams of 7.25 MeV and 5 MeV (energy spread approximately 20%) along the same axis and observing the respective NRF gamma rays of the isotopes, it is possible to measure elemental ratios. By finding the above-mentioned elemental ratios, it is possible to identify the explosive's type. By finding the abundances, it is possible to find the amount. With respect to detecting specific isotopes (e.g., carbon-12, nitrogen-14, etc.), the energy of the electron beam by the electron accelerator, the wavelength from the laser sources, laser polarization by the wave plates, and the radiation detectors are, by way of example, combined in the following manner. Electron energy: 641 MeV Laser polarization: Vertical 4439 keV of 12C (E2) (laser: wavelength 1550 nm, detector: perpendicular direction) 4915 keV of 14N (E1) (laser: wavelength 1550 nm, detector: horizontal direction) 6917 keV of 16O (E2) (laser: wavelength 1064 nm, detector: perpendicular direction) 7029 keV of 14N (M1) (laser: wavelength 1064 nm, detector: perpendicular direction) 7117 keV of 16O (E1) (laser: wavelength 1064 nm, detector: horizontal direction) It is thus possible to identify the five NRF levels (three nuclides) above. Alternatively, by combining the energy of the electron beam, wavelength from the laser sources, the laser polarization by the wave plates, and the radiation detectors as follows, it is possible to identify two NRF levels (two nuclides). Electron energy: 530 MeV Laser polarization: Vertical 4439 keV of 12C (E2) (laser: wavelength 1064 nm, detector: perpendicular direction) 4915 keV of 14N (E1) (laser: wavelength 1064 nm, detector: horizontal direction) 1 Photon beam 2 Sample 3 Isotope 4 NRF gamma ray 5 Scatter X-ray 6 High energy resolution radiation detector 7 Photon intensity monitor 11 Electron beam accelerator 12 Electron beam 13 Laser light source 14 Wave plate 15 Mirror 16 Laser light 17 Laser light source 18 Wave plate 19 Mirror 20 Laser light 21 Pulsed LCS photon beam 22 Pulsed LCS photon beam 23 Collimator 31 Sample 32 Radiation detector 33 Radiation detector 34 Photon intensity monitor 35 Movable plate 36 Drive part 37 Control processing part 38 Monitor 51 LCS photon beam 52 LCS photon beam 53 NRF gamma ray 54 NRF gamma ray 61 Sample 64 Radiation detector 65 Photon intensity monitor 66 Radiation detector 71 Shield box 72 Concealed object 73 Detector 74 NRF gamma ray 75 NRF gamma ray |
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055442064 | summary | FIELD OF THE INVENTION The present invention is directed to the field of nuclear reactor inspection and repair. In particular, the present invention is directed to an apparatus for inspecting nuclear reactors and power stations. BACKGROUND OF THE INVENTION In a nuclear reactor power plant, a nuclear reactor vessel is used to generate heat for the production of steam and electricity. The reactor vessel is typically a pressure vessel which encloses a core of nuclear fuel and cooling water which is often borated water. In order to monitor the operating conditions within the pressure vessel, instrumentation devices are often introduced into the nuclear core through ports or penetrations in the vessel. Some of these penetrations are provided through the reactor vessel closure head. The closure head may also provide for the penetration of drive mechanisms of control rods, which are used to regulate the rate of nuclear reactions which take place within the core, and which control the power output of the plant. Although these ports are mechanically sealed to prevent the inadvertent leakage of coolant from the reactor vessel, the operating conditions of a nuclear reactor pressure vessel require additional safeguards. Typically, the pressure vessel maintains the coolant therein at an internal pressure of about 15 MPa (2250 psi) and at a temperature of about 315.degree. C. (600.degree. F.). Because of such an enlarged internal pressure, reactor coolant may leak from the mechanical joint of these penetrations, or when the control rods are withdrawn. The coolant within a reactor vessel is typically acidic and highly corrosive due to the presence of boric acid which is dissolved within the coolant. Boric acid is a neutron absorber which is used as a variable reactivity control over the long-term operation of the plant. Even though there are regulatory limits on the allowable amounts of coolant which may be emitted from the reactor vessel, components on the exterior of and in close proximity to, the reactor vessel head need to be periodically inspected to determine if coolant is being emitted. Because an operating nuclear reactor generates an irradiated environment, the inspection and/or maintenance of the reactor vessel is typically conducted at times when the reactor is shut down for normal inspection or maintenance procedures, such as, the refueling of the core. A usual tell-tale sign of the presence of a leak in this area is white boric acid crystal deposits on the reactor vessel head. Any of several methods for determining the presence of a possible leak source may be used, but verification of the existence of a leak, estimation of its size, and the identification of its location is best done visually. Generally, control rod drive mechanisms and instrumentation ports are enclosed by a cooling shroud. The shroud provides protection for the drive mechanisms, as well as a means for directing the flow of air around the ports for natural circulation, thereby cooling of the ports and drive mechanisms. This can make it even more difficult to visually detect the presence of borated coolant in this area as well as perforations in the shroud of the reactor. A number of patents are directed to nuclear power plant inspection and inspection devices and in particular, devices for inspecting nuclear reactors. U.S. Pat. No. 4,857,261 to Marshall discloses a monitoring system for use in a nuclear reactor head area. The monitoring system disclosed in the Marshall patent comprises a plurality of video cameras attached to a cooling shroud on the vessel head. A right-angle lens is attached to each video camera and a halogen light source is attached to the shroud adjacent to each video camera to provide adequate lighting. U.S. Pat. No. 3,780,571 to Wiesener discloses a device for inspecting nuclear reactor pressure vessels, which can be removed from the vessel when the vessel is in service and later replaced in the same position to make inspections. The device includes means for handling inspection of equipment, such as cameras, ultrasonic detectors or the like, and is moveable in several motions to permit it to operate in all areas of the vessel. U.S. Pat. No. 4,311,556 to Iwamoto discloses a system for the inspection of the inside of a nuclear reactor vessel which utilizes a working unit and monitor for monitoring the movement of the working unit by a manipulator. The manipulator is a 2-arm structure, one of which is equipped with a working unit and the other of which is equipped with a television camera to continuously image the condition of the working unit. U.S. Pat. No. 4,460,920 discloses an automatic traveling tube-interior manipulator for the remotely controlled transportation of testing devices and tools along given feed paths. The manipulator has a stepping mechanism that enables it to traverse narrow tube elbows or even vertical tube sections so that the test device tools attached to the manipulator head can be positioned and moved along their feed paths. U.S. Pat. No. 4,643,867 to Hornak, et al. discloses a mounting system for a set of four television cameras upon a nuclear reactor refueling machine outer or stationery mast. The television cameras scan the reactor core fuel assemblies as they are being vertically removed from the core by the refueling machine during the performance of the refueling operation. In the alternative, the television cameras provide remote viewing of the reactor core in order to facilitate insertion of the fuel assemblies into the core during a refueling operation. U.S. Pat. No. 4,656,509 to Matsuyama discloses a water leakage monitoring system for remotely monitoring leakages of water from monitored members and equipment disposed in a plant. The drive of the Matsuyama patent comprises a remotely controlled carriage which intermittently travels along an inspection route, a television camera hung from the carriage so as to monitor water leakage through each of the monitored members along the inspection path through the operation of an attitude adjusting mechanism in response to an attitude control signal. A coating composition applied to the surface of each of the monitored members changes in color when it is wet with water leaked through the associated member. U.S. Pat. No. 4,661,308 to Tokinaka discloses a remote-controlled mobile inspecting and monitoring system which comprises a rail laid along an inspection route, power lines provided along the rail, and an inspection vehicle carrying inspecting and monitoring equipment to move along the rail. A power receiving unit mounted on the inspection vehicle is connected to the power lines such that the inspection vehicle carries a battery unit for supplying power to on board equipment and a change over unit for controlling the power receiving unit so as to be connected to or disconnected from the power lines. The system is capable of reducing the load on the battery unit by receiving power through the power receiving unit from the power lines so that a battery unit of reduced capacity is needed, thereby making it possible to reduce the size of the inspection vehicle. U.S. Pat. No. 4,668,981 to Egger discloses a system for performing visual inspection of water-immersed portions of a nuclear reactor vessel. The system utilizes the manual manipulation of a television camera through which the reactor components are viewed. A television camera is suspended within the reactor vessel from a first hand held line by an individual positioned on a platform above the open top of the vessel. The camera is carried on a positioning device which includes a pair of arms extending outwardly to terminal ends on opposite sides of the camera axis. Lines attached to each of these arms are held by a second individual, also positioned on the platform, at a distance from the first individual. A third individual is positioned at a television monitor to view the image relayed by the camera, and to direct the other two individuals in its deployment. U.S. Pat. No. 4,702,878 to Klug, et al, discloses an apparatus for remotely inspecting and cleaning areas between the tubes on a tube sheet in a nuclear steam generator. A flexible tube is connected to one end of the sled and has at its other end extending out through an opening. The tube is of sufficient length to be freely movable into and out of the shell opening to move the sled along the periphery of the tube sheet. A probe such as a television camera or a fiberscope and a gripper are mounted on the sled for searching and retrieving objects. Finally, U.S. Pat. No. 4,729,423 to Martin discloses a process and apparatus for optically checking the shape and dimensions of the ends of the tubes of a steam generator. A photographic camera is placed parallel to the tubular plate inside the water tank and exposures are made with the camera in different positions so as to obtain at least two views at different angles of the tubes which have to be checked. None of the prior art discussed above discloses a nuclear reactor inspection device which can fully inspect a reactor head and which can be used in association with a plurality of movable members which permit the complete visual inspection of the reactor and reactor head. It would be desirable to provide a nuclear reactor inspection device which can easily inspect the reactor head of a nuclear reactor and which can be utilized to inspect selected portions of a reactor head including the control rod drive mechanism penetrations, the thermal sleeves, the funnels, and the annulus between the control rod drive mechanism and the thermal sleeves. It would be particularly desirable to provide a nuclear reactor head inspection system which can rotate a full 360.degree. and which can provide for full vertical and horizontal visibility and movement of a camera and light assembly. The present invention is directed to overcome the problems associated with prior art devices for inspecting the reactor head of a nuclear power plant. The objects and benefits of the present invention will become apparent from the detailed description and summary which follow. SUMMARY OF THE INVENTION In accordance with the present invention, apparatus for inspecting a nuclear reactor is disclosed. The invention comprises: a boom rotatably connected to a pivot point on a base; means for controllably rotating the boom with respect to said pivot point on said base; and a camera assembly affixed to said boom and being longitudinally slidable thereto such that the camera assembly may be slidably adjusted with respect to said boom. In accordance with a more preferred embodiment, the present invention is directed to apparatus for inspecting the reactor head of a nuclear reactor comprising: a boom rotatably connected to a base at a pivot point, said base being affixed atop a nuclear reactor, said boom having a caster at one end such that said boom can rotate with respect to said pivot point and said caster; means for selectively rotating and positioning said boom and said caster about said pivot point; a camera slidably mated with said boom between first and second positions on said boom; means for slidably moving said camera between said first and second positions along said boom; means for rotating said camera means in a first direction with respect to said boom; and means for rotating said camera means in a second direction with respect to said boom. In yet a further embodiment, the present invention is directed to apparatus for inspecting the reactor head of a nuclear reactor is disclosed comprising: a boom rotatably fixed about a pivot point on said reactor head, said boom having a caster at one end such that said boom can rotate 360.degree. with respect to said pivot point and said caster; means for rotating said boom about said caster; camera and light means slidably mated with said boom such that said camera and light means can move between first and second positions on said boom; and means for moving said camera and light means between said first and second positions on said boom. These and other advantages of the present invention will become apparent from the detailed description and claims which follow. |
abstract | Disclosed are a fuel rod and a fuel bundle using the fuel rod. The fuel rod may include first enriched uranium in a boost zone of the fuel rod, wherein the boost zone may be arranged directly at a bottom of the fuel rod. The fuel rod may also include second enriched uranium in a second zone of the fuel rod, wherein the second zone is arranged over the boost zone. The fuel rod may also include natural uranium in a third zone of the fuel rod, wherein the third zone is arranged over the second zone. In this fuel rod, a percent of enrichment of the enriched uranium in the boost zone is at least one percent. |
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description | This application is based upon and claims the benefit of priority from the prior Japanese Patent Applications No. 2008-274689, filed on Oct. 24, 2008, and No. 2009-242869, filed on Oct. 21, 2009; the entire contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to an extreme ultraviolet light source apparatus generating plasma by irradiating a target with a laser light and outputting ultraviolet light emitted from the plasma. 2. Description of the Related Art In recent years, along with a progress in miniaturization of semiconductor device, miniaturization of transcription pattern used in photolithography in a semiconductor process has developed rapidly. In the next generation, microfabrication to the extent of 70 nm to 45 nm, or even to the extent of 32 nm and beyond will be required. Therefore, in order to comply with the demand of microfabrication to the extent of 32 nm and beyond, development of such exposure apparatus combining an extreme ultraviolet (EUV) light source for a wavelength of about 13 nm and a reflection-type reduction projection optical system is expected. As the EUV light source, there are three possible types, which are a laser produced plasma (LPP) light source using plasma generated by irradiating a target with a laser beam, a discharge produced plasma (DPP) light source using plasma generated by electrical discharge, and a synchrotron radiation (SR) light source using orbital radiant light. Among these light sources, the LPP light source has such advantages that luminance can be made extremely high as close to the black-body radiation because plasma density can be made higher. Moreover, the LPP light source also has an advantage that luminescence only with a desired wavelength band is possible by selecting a target material. Furthermore, the LPP light source has such advantages that there is no construction such as electrode around a light source because the light source is a point light source with nearly isotropic angular distributions, extremely wide collecting solid angle can be acquired, and so on. Accordingly, the LPP light source having such advantages is expected as a light source for EUV lithography which requires more than several dozen to several hundred watt power. In the EUV light source apparatus with the LPP system, firstly, a target material supplied inside a vacuum chamber is excited by being irradiated with a laser light and thus be ionized to become plasma. Then, a cocktail light with various wavelength components including an EUV light is emitted from the generated plasma. Then, the EUV light source apparatus focuses the EUV light by reflecting the EUV light using an EUV collector mirror which selectively reflects an EUV light with a desired wavelength, e.g. a 13.5 nm wavelength component. The reflected EUV light is inputted to an exposure apparatus. On a reflective surface of the EUV collector mirror, a multilayer coating with a structure in that thin coating of molybdenum (Mo) and thin coating of silicon (Si) are alternately stacked, for instance, is formed. The multilayer coating exhibits a high reflectance ratio (of about 60% to 70%) with respect to the EUV light with a 13.5 nm wavelength. Here, as mentioned above, plasma is generated by irradiating a target with a laser light, and at the same time, particles (debris) such as gaseous ion particles and neutral particles, and tiny particles (metal cluster) which have not been able to become plasma fly out around thereof from a plasma luminescence point. The debris fly toward surfaces of various optical elements such as an EUV collector mirror located in the vacuum chamber, focusing mirrors or focusing lenses for focusing a laser light on a target, and other optical system for measuring an EUV light intensity, and so forth. Therefore, fast ion debris with comparatively high energy erode surfaces of optical elements and damage reflective coating and non-reflective coating of the surfaces. As a result, the surfaces of the optical elements will become a metal component, which is a target material. On the other hand, slow ion debris with comparatively low energy and neutral particle debris will deposit on surfaces of optical elements. As a result, a layer of a compound of metal, which is a target material, is formed on the surfaces of the optical elements. As a result of the debris entering as mentioned above, the reflective coating and the non-reflective coating of each optical element is damaged or a compound layer is formed on the surfaces of the optical elements, whereby reflectance or transmittance of the optical elements decrease and the optical elements become unusable. In this respect, Japanese patent application Laid-Open No. 2005-197456 discloses a technique such that debris flying from plasma is trapped by a magnetic field generated inside an optical collecting system by a magnetic field generator when current is supplied to the magnetic field generator. According to this technique, by locating a luminescence point of an EUV light within the magnetic field, ion debris flying from the plasma generated around the luminescence point converge in a direction of the magnetic field by Lorentz force by the magnetic field. As a result, contamination of neighboring optical elements with debris and damages of the optical elements can be reduced. On the other hand, US patent application Laid-Open No. 2008/0197297 discloses a technique with which generated debris are trapped around a plasma luminescence point by a magnetic field generated as a result of making the plasma luminescence point surrounded with wiring and passing current to the wiring. In accordance with one aspect of the present invention, an extreme ultraviolet light source apparatus comprising a vacuum chamber, a target supply unit supplying a target into the vacuum chamber, a laser oscillator outputting a laser light into the vacuum chamber, and a collector mirror outputting an extreme ultraviolet light outside by reflecting the extreme ultraviolet light emitted from the target being ionized as a plasma by irradiation with the laser light at a plasma luminescence point in the vacuum chamber, the extreme ultraviolet light source apparatus comprises an ion debris removal unit at least a part of which is located in an obscuration region including the plasma luminescence point. In accordance with another aspect of the present invention, an extreme ultraviolet light source apparatus comprising a vacuum chamber having an optical window for inputting a laser light inside from outside, a target supply unit supplying a target into the vacuum chamber, and a collector mirror outputting an extreme ultraviolet light outside by reflecting the extreme ultraviolet light emitted from the target being ionized as a plasma by irradiation with the laser light at a plasma luminescence point in the vacuum chamber, the extreme ultraviolet light source apparatus comprises an ion debris removal unit at least a part of which is located in an obscuration region including the plasma luminescence point. These and other objects, features, aspects, and advantages of the present invention will become apparent to those skilled in the art from the following detailed description, which, taken in conjunction with the annexed drawings, discloses preferred embodiments of the present invention. Now, exemplary embodiments of an extreme ultraviolet light source apparatus according to the present invention will be described in detail with reference to the accompanying drawings. FIG. 1 is a schematic diagram showing a structure of an extreme ultraviolet light source apparatus according to a first embodiment of the present invention. FIG. 2 is a schematic diagram showing a structure of an ion debris removal unit in the extreme ultraviolet light source apparatus shown in FIG. 1. In FIGS. 1 and 2, the extreme ultraviolet light source apparatus has a vacuum chamber 1 in which a plasma luminescence point P1 is located. In this particular embodiment, the vacuum chamber may be a low pressure chamber that includes a small amount of gas other than a high vacuum chamber. Moreover, for instance, it is possible that the vacuum chamber contains a buffer gas (He gas, Ar gas, or the like) and an etching gas, of which pressure is to the extent that does not influence the operation of the ion debris removal unit. In particular, when a target is Sn, as for the etching gas, low pressure hydrogen gas, halogen gas, halogen hydride gas, hydrogen radical gas, or the like can be used. These etching gases have high transmittance with respect to an EUV light while being able to etch Sn. A CO2 pulse laser light La emitted from a drive laser 2 which is located outside the vacuum chamber 1 enters inside the vacuum chamber 1 via a window 1a provided for inputting a laser light of the vacuum chamber 1. The CO2 pulse laser light La is focused on the plasma luminescence point P1 through an optical collecting system 3 within the vacuum chamber 1 and an aperture 4a of an EUV collector mirror 4. A molten Sn tank 5 stores a liquid metal of Sn, and discharges droplets 7 of Sn as being targets from a nozzle 6. The nozzle 6 discharges the droplets 7 so that the droplets 7 pass through the plasma luminescence point P1. Here, in order to irradiate each droplet 7 with the CO2 pulse laser light La at the plasma luminescence point P1, each output timing of the droplet 7 and each pulse timing of the CO2 pulse laser light La are controlled to be synchronous by a controller (not shown). In the vacuum chamber 1, the EUV collector mirror 4 is mounted. The EUV collector mirror 4 reflects an EUV light Lb emitted from the plasma generated by the droplet 7 being irradiated with the CO2 pulse laser light La, in such a way that the EUV light Lb is focused. The reflected EUV light Lb is outputted from a window of a vacuum gate valve 1b in the vacuum chamber to an EUV exposure apparatus 11. Here, in the vacuum chamber 1, an ion debris removal unit 10 for trapping charged particles generated from the plasma produced at the plasma luminescence point P1 is mounted. The ion debris removal unit 10 is mounted so that at least a part of the ion debris removal unit 10 is located in an obscuration region E. The obscuration region E refers to a region corresponding to an angular range in which EUV light Lb focused by the EUV collector mirror 4 will not be used in the EUV exposure apparatus 11. That is, the EUV light Lb emitted from the plasma luminescence point P1 is focused on a focus position P2 by the EUV collector mirror 4. In this explanation, such three-dimensional solid region at the focus point P2 corresponding to the angular range in which the EUV light Lb will not be used in the EUV exposure apparatus 11 is defined as the obscuration region E. Usually, the EUV light in the obscuration region E is not used in the exposure apparatus 11. Therefore, exposure performance and throughput of the exposure apparatus will not be influenced even if the EUV light in the obscuration region E is not inputted to the exposure apparatus. The obscuration region E is spatially distributed in each of both of the extreme ultraviolet light source apparatus and the EUV exposure apparatus. In this respect, in the first embodiment, the ion debris removal unit 10 is arranged inside the obscuration region E. Thereby, it is possible to mount the ion debris removal unit 10 inside the extreme ultraviolet light source apparatus without reducing power of the EUV light that is used for exposure in the EUV exposure apparatus 11. Moreover, the extreme ultraviolet light source apparatus has a droplet collection cylinder 8 which collects the droplets 7 outputted from the nozzle 6. The droplet collection cylinder 8 is located at a place facing the nozzle 6. Residual droplets such as droplets not used for generation of plasma and droplets remaining after the generation of plasma are collected by the droplet collection cylinder 8. The collected droplets can be reused as droplets by being supplied to the molten Sn tank 5 again after a recycling process. A dumper 12 absorbs the CO2 pulse laser light La which was not emitted to the droplet 7 or which did not contribute to generate plasma even if the CO2 pulse laser light La was emitted to the droplet 7. Moreover, an adjustment camera 14 and an EUV light meter 13 monitor a position of the droplet 7 and developmental state of plasma in the plasma luminescence point P1 in order to control output timing of each droplet 7 and output timing of the CO2 pulse laser light La. Furthermore, a reflection and collection EUV light meter 15 measures a power of the EUV light Lb outputted from the extreme ultraviolet light source apparatus based on a part of the EUV light Lb reflected by a spectrum pass filter SPF. Furthermore, a vacuum evacuation apparatus 16 is a turbo molecular pump (TMP), or the like, and maintains degree of vacuum in the vacuum chamber 1. As shown in FIG. 2, the ion debris removal unit 10 has a pair of electromagnetic coils 21 and 22 mounted in such a way as to sandwich the plasma luminescence point P1. As described above, the pair of electromagnetic coils 21 and 22 are located in the obscuration region E. By passing current to the pair of electromagnetic coils 21 and 22 as shown in FIG. 2, a magnetic field of which magnetic direction is a direction passing through inside both bores of the electromagnetic coils 21 and 22, is generated. The magnetic field includes a convergence region expanding in the magnetic direction around the plasma luminescence point P 1. The ion debris such as Sn+ are converged inside the convergence region. The trapped ion debris adheres to inner walls of ion collection (receiving) cylinders 23 and 24 mounted outside the electromagnetic coils 21 and 22, respectively, after passing through the bores of the electromagnetic coils 21 and 22. The ion debris having adhered to the inner walls of the ion collection (receiving) cylinders 23 and 24 is collected later. In FIG. 2, the ion collection (receiving) cylinders 23 and 24 are mounted outside the obscuration region E, but it is also possible that the ion collection (receiving) cylinders 23 and 24 are mounted inside the obscuration region E. The pair of the electromagnetic coils 21 and 22 can be located near the plasma luminescence point P1 as being closely arranged with a several mm distance in between, for instance. Therefore, even if downsized electromagnetic coils are used as the electromagnetic coils 21 and 22, it is possible to generate a local magnetic field with an approximately several T near the plasma luminescence point P1. Furthermore, due to the electromagnetic coils 21 and 22 are located inside the obscuration region E, the EUV light Lb will not be shadowed. Moreover, by using the downsized electromagnetic coils 21 and 22, it is possible to downsize an occupation space in the vacuum chamber 1, and therefore, it is possible to enhance miniaturization of the extreme ultraviolet light source apparatus. Moreover, when the downsized electromagnetic coils 21 and 22 are used, a density of magnetic flux formed outside the plasma luminescence point P1 is low, meaning that is a leak magnetic field from the extreme ultraviolet light source apparatus can be attenuated, and therefore, it is possible to reduce possible influence of the leak magnetic field on apparatuses other than the extreme ultraviolet light source apparatus, e.g. the EUV exposure apparatus 11. Next, a second embodiment of the present invention will be described in detail. In the above-described first embodiment, the ion debris removal unit 10 forms the local magnetic field near the plasma luminescence point P1 in order to trap and collect ion debris. On the other hand, in the second embodiment, a local electrical field is formed near the plasma luminescence point P1 in order to trap and collect ion debris. FIG. 3 is a vertical cross-sectional diagram showing a structure of an ion debris removal unit in an extreme ultraviolet light source apparatus according to the second embodiment of the present invention. As shown in FIG. 3, the ion debris removal unit 20 is located inside the obscuration region E. On a side of the EUV collector mirror 4 with respect to the plasma luminescence point P1, a centroclinal electrostatic grid 30 is mounted. Moreover, on an opposite side of the EUV collector mirror 4 with respect to the plasma luminescence point P1, a cylindrical ion collection (receiving) cylinder 33 is mounted while a perforated circular disk 31 is sandwiched between the plasma luminescence point P1 and the ion collection (receiving) cylinder 33. As shown in FIG. 4, the electrostatic grid 30 and the perforated circular disk 31 are electrically connected with each other and have a positive electric potential of about 1 to 3 kV applied to, respectively. On the other hand, the ion collection (receiving) cylinder 33 is grounded. The electrostatic grid 30 is a grid with an aperture ratio of over 90%, and does not substantially disturb incidence of the CO2 pulse laser light La or emission of the EUV light Lb. Furthermore, at a center of the perforated circular disk 31, an aperture of 10 mm in diameter is opened. As described above, a size of the ion debris removal unit 20 is small enough to fit within the obscuration region E. As the size of the ion debris removal unit 20, a diameter thereof is about 30 mm, for instance. More preferably, as shown in FIGS. 5 and 6, the electrostatic grid 30 has an about 1 kV negative voltage applied to, while another grid 35 is mounted inside the electrostatic grid 30 in order to rebound approximately the same number of electrons that are generated along with ions. Thereby, it is possible to remove both ions and electrons. Positive charged ion debris charged by the positive electric potential at the electrostatic grid 30 is rebounded by a reactive Coulombic force and gravitated toward a side of the ion collection (receiving) cylinder 33, which is a lower electric potential side, through the perforated circular disk 31. In addition, in order to form a gap, an isolator 32 with electrical resistance such as AL2O3 is mounted between each the electrostatic grid 30 and the ion collection (receiving) cylinder 31, and the perforated circular disk 31 and the ion collection (receiving) cylinder 33. It is preferred that an interval of the gap is good enough to not cause insulation breakdown, e.g. over 5 mm, for instance. In the ion collection (receiving) cylinder 33, a conular projection 34 with a vertex facing toward the EUV collector mirror 4 is mounted. Due to the vertex of the projection 34 facing an incident side of the CO2 pulse laser light La, a surface area of a surface irradiated with the CO2 pulse laser light La becomes larger, and thereby, it is possible to improve a dumper performance with respect to the CO2 pulse laser light La. The ion debris will adhere to an inner wall of the ion collection (receiving) cylinder 33, and then collected. As for the perforated circular disk 31, a plate-like SiC or MN with artificial diamond coating on a surface is used. However, it is not limited to such arrangement, while the perforated circular disk 31 can be made from a material with heat resistance and high conductivity. Furthermore, it is preferable that the whole ion collection (receiving) cylinder 33 is thermally controlled to be over the melting point of the target material (230 degrees C. being the melting point of the Sn, for instance) in order to liquefy and discharge the collected (received) ion debris. Furthermore, it is preferable that the ion collection (receiving) cylinder 33 is made from Cu or the like, which is a material with high conductivity. Moreover, it is preferable that a surface of the ion collection (receiving) cylinder 33 is coated with Mo, C and Ti, or the like, which show high resistant against ion-sputtering. Moreover, when the surface of the ion collection (receiving) cylinder 33 is coated with Mo as being a component material of a multilayer coating of the EUV collector mirror, it is possible to prevent the reflectance of the EUV collector mirror from decreasing even if the surface of the ion collection (receiving) cylinder 33 is sputtered. In the second embodiment, by having the structure in that ion debris is collected (received) by having the local magnetic field formed in the obscuration region E, it is possible to obtain the same effects as in the first embodiment. The other structures and effects are the same as in the above-described embodiment, and redundant explanations will be omitted herein. Next, a third embodiment of the present invention will be described in detail. In the third embodiment, in addition to the ion debris removal units 10 and 20, members for supporting the ion debris removal units 10 and 20, respectively, are mounted in the obscuration region E. FIG. 7 is a horizontal cross-sectional diagram showing a structure of an extreme ultraviolet light source apparatus according to the third embodiment of the present invention. FIG. 8 is a vertical cross-sectional diagram showing a structure of the extreme ultraviolet light source apparatus according to the third embodiment of the present invention. As shown in FIGS. 7 and 8, a cross-section perpendicular to an optical axis of the EUV light Lb passing through a region (hereinafter to be referred to as an EUV area Eb) where the EUV light Lb reflected by the EUV collector mirror 4 passes through may include not only a central region but also a radial obscuration regions E which expand radially from the central region. In this particular embodiment, supporting members 40 for supporting the ion debris removal unit 10 are arranged in such radial obscuration region E. The supporting members 40 are fixed on a wall surface of the vacuum chamber 1. Furthermore, in the third embodiment, as shown in FIG. 9, it is possible to support the ion collection (receiving) cylinder 50 using a collection tube 51 in place of the supporting members 40. The ion collection (receiving) cylinder 50 and the collection tube 51 are covered with a heater 52 which is a temperature regulator, for instance, and thermally controlled to a temperature at which Sn being the ion debris in the ion collection (receiving) cylinder 50 and the collection tube 51 will melt. The molten ion debris are disembogued outside the vacuum chamber 1 via the collection tube 51. In this case, it is preferable to have an axis of the collection tube 51 face in a gravity direction GA such that the molten Sn can be easily discharged outside the vacuum chamber 1. Here, depending on the arrangement of the obscuration region E, there may be a case in that the axis of the collection tube 51 can not be made to face in the gravity direction GA. Even in such case, as shown in FIG. 10, it is preferable to tilt the axis of the collection tube 51 to the gravity direction GA within the obscuration region E in such a way that the collection tube 51 does not protrude from the obscuration region E. The other structures and effects are the same as in the above-described embodiments, and redundant explanations will be omitted herein. Next, a fourth embodiment of the present invention will be described in detail. In the above-described first to third embodiments, ion debris is collected (received) by having a local magnetic field or local electric field formed using the ion debris removal unit 10 or 20. On the other hand, in the fourth embodiment, the ion debris removal unit 10 forms an accessorial local electric field in addition to the local magnetic field. Thereby, the collection (receiving) rate and collection (receiving) efficiency of ion debris can be further improved, and thus it is possible to reduce possible collision of debris directed to the EUV collector mirror 4. FIG. 11 is a schematic diagram showing an outline structure of an extreme ultraviolet light source apparatus according to the fourth embodiment of the present invention. As shown in FIG. 11, in the extreme ultraviolet light source apparatus according to this particular embodiment, as in the first embodiment, a pair of small electric magnets 61 and 62 are located in the obscuration region E. The small electric magnets 61 and 62 form a convergence region E2 where ion debris generated at the plasma luminescence point P1 are converged in bore directions (magnetic field direction) and trapped. The ion debris (positive ion debris, negative ion debris) trapped in the convergence region E2 will drift depending on the polar character thereof, respectively, and then, is collected (received) by the ion collection (receiving) cylinders 65 and 66. Furthermore, the extreme ultraviolet light source apparatus according to the fourth embodiment comprises an electrostatic grid 63 having a concave face with a solid angle covering the EUV collector mirror 4 with respect to the plasma luminescence point P1, and a ground electrode 64 located in the obscuration region E as facing the electrostatic grid 63 through the plasma luminescence point P1. The electrostatic grid 63 is located on a side of the EUV collector mirror 4 with respect to the plasma luminescence point P1 while being outside the obscuration region E, and functions as the electrostatic grid 30 in the second embodiment. By this arrangement, in this particular embodiment, it is possible to draw back high energy ion debris flying toward the EUV collector mirror 4 to the side of the plasma luminescence point P1 by a reactive Coulombic force. In the fourth embodiment, as described above, the local electric field that draws back the high energy ion debris flying out from the local magnetic field to the plasma luminescence point P1 is formed. As a result, according to this particular embodiment, it is possible to reduce influence of ion debris on the desired optical elements while it is possible to improve the collection (receiving) efficiency of ion debris. The other structures and effects are the same as in the above-described embodiments, and redundant explanations will be omitted herein. Next, a fifth embodiment of the present invention will be described in detail. FIG. 12 is a horizontal cross-sectional diagram showing an outline structure of an extreme ultraviolet light source apparatus according to the fifth embodiment of the present invention. FIG. 13 is a vertical cross-sectional diagram showing an outline structure of the extreme ultraviolet light source apparatus according to the fifth embodiment. FIG. 14 is a diagram showing an example of a far field pattern formed on A-A surface in FIG. 12. As shown in FIGS. 12 to 14, the extreme ultraviolet light source apparatus according to this particular embodiment has magnetic cores 21a and 22a respectively extending from the electromagnetic coils 21 and 22 which are located outside the passing region (the EUV area Eb) of the EUV light Lb reflected by the EUV collector mirror 4. Ion debris generated around the plasma luminescence point P1 are converged in a certain definite range (the convergence region E2) by a magnetic field formed by the magnetic cores 21a and 22a when electrical current is applied to the electromagnetic coils 21 and 22. The magnetic cores 21a and 22a extend from the electromagnetic coils 21 and 22, respectively, into the obscuration region E in the EUV area Eb. Here, as a far field pattern in FIG. 14 indicates, the EUV area Eb includes an obscuration region Ec corresponding to the aperture 4a at the center portion, and further includes strip-shaped obscuration regions EDa and EDb expanding radially from an obscuration region Ec located at the center portion to an exterior edge of the EUV area Eb. The strip-shape obscuration regions EDa and EDb are regions which are shadowed by the optical systems located in the vacuum chamber 1 and by ion debris drifted from the plasma luminescence point P1. Therefore, in this particular embodiment, the magnetic cores 21a and 22a are located in such a way that the parts extending into the EUV area Eb are located as being included in the strip-shape obscuration regions EDa and EDb, respectively. Accordingly, it is possible to locate the magnetic cores 21a and 22a so that heads of the magnetic cores 21a and 22a as being origins of the magnetic field face closely to each other while sandwiching the plasma luminescence point P1 in between without shutting off the EUV light Eb to be used for exposure in the EUV exposure apparatus 11. As a result, it is possible to locally form a strong magnetic field around the plasma luminescence point P1, and it is possible to steadily converge ion debris generated around the plasma luminescence point P1 and effectively lead the ion debris toward desired directions (directions toward the ion collection (receiving) cylinders 23 and 24). In this particular embodiment, as shown in FIGS. 12 and 14, at least a part of or whole of the ion collection (receiving) cylinders 23 and 24 are located in the bores of the electromagnetic coils 21 and 22, respectively. Specifically, the ion collection (receiving) cylinders 23 and 24 are located inside the tubular magnetic cores 21a and 22a which are arranged as passing through the insides of the bores of the electromagnetic coils 21 and 22. Thereby, it is possible to have the ion collection (receiving) cylinders 23 and 24 steadily collect (receive) ion debris drifting along the magnetic field having been formed by the magnetic cores 21a and 22a and passing through the bore of each of the electromagnetic coils 21 and 22. In this arrangement, by locating at least parts of the ion collection (receiving) cylinders 23 and 24 inside the obscuration regions EDa and EDb in the EUV area Eb, it is possible to shorten a distance between the plasma luminescence point P1 and the ion collection (receiving) cylinders 23 and 24, respectively. Thus, it is possible to make a trapping solid angle with respect to an existence area of ion debris (the convergence region E2) larger, whereby it is possible to have the ion collection (receiving) cylinders 23 and 24 steadily collect (receive) ion debris generated around the plasma luminescence point P1. In this respect, in this particular embodiment, the ion collection (receiving) cylinders 23 and 24 are located in the strip-shape obscuration regions EDa and EDb that expand from the obscuration region Ec to the exterior edge of the EUV area Eb. By this arrangement, as compared with the case where the ion collection (receiving) cylinders 23 and 24 are located in the obscuration region Ec, required degree of miniaturization for the ion collection (receiving) cylinders 23 and 24 can be relaxed. As a result, it is possible to make the ion collection (receiving) cylinders 23 and 24 retaining a sufficient trapping solid angle with respect to the convergence region E2, whereby it is possible to have the ion collection (receiving) cylinders 23 and 24 steadily collect (receive) ion debris drifting from around the plasma luminescence point P1. Moreover, by locating the ion collection (receiving) cylinders 23 and 24 in the obscuration regions EDa and EDb, it is possible to space between the plasma luminescence point P1 and the ion collection (receiving) cylinders 23 and 24, respectively, to a certain degree, and therefore, it is also possible to achieve an effect that possible damages of the ion collection (receiving) cylinder 23 and 24 caused by the high energy ion debris generated around the plasma luminescence point P1 can be reduced. The other structures and effects are the same as in the above-described embodiments, and redundant explanations will be omitted herein. Furthermore, in this particular embodiment, the case where the electromagnetic coils 21 and 22 are located outside the EUV area Eb but inside the vacuum chamber 1 has been described as an example. However, the present invention is not limited to such arrangement. The electromagnetic coils 21 and 22 can also be located outside the vacuum chamber 1 or inside the obscuration region E, for instance. Next, a sixth embodiment of the present invention will be described in detail. FIG. 15 is a diagram showing an electromagnetic coil and an ion collection (receiving) cylinder in an extreme ultraviolet light source apparatus according to the sixth embodiment of the present invention. As shown in FIG. 15, the extreme ultraviolet light source apparatus according to this particular embodiment has the same structure as the extreme ultraviolet light source apparatus according to the above-described fifth embodiment, except that surfaces of the magnetic cores 21a and 22a are coated with coating layers 21b and 22b, respectively. The coating layers 21b and 22b are coatings for preventing the surfaces of the magnetic cores 21a and 22a from being sputtered by ion debris. In order to form a strong magnetic field around the plasma luminescence point P1, it is effective that the magnetic cores 21a and 22a are arranged to be located around the plasma luminescence point P1 as close as possible. However, if the magnetic cores 21a and 22a are too close to the plasma luminescence point P1, the magnetic cores 21a and 22a may be sputtered by comparatively high energy ion debris generated around the plasma luminescence point P1. Particles made from a construction material of the magnetic cores 21a and 22a, which are generated by the magnetic cores 21a and 22a being sputtered, adhere to the EUV collector mirror 4, and so on, to become a factor that decreases reflectance. Therefore, it is preferable to prevent the magnetic cores 21a and 22a from being sputtered wherever possible. Accordingly, as already mentioned, in this particular embodiment, the magnetic cores 21a and 22a are coated with the coating layers 21b and 22b. Thus, it is possible to prevent the magnetic cores 21a and 22a from being sputtered, whereby it is possible to locate the magnetic cores 21a and 22a near the plasma luminescence point P1 more closely. As a result, it is possible to steadily and effectively lead ion debris to the ion collection (receiving) cylinders 23 and 24. As a material of the coating layers 21b and 22b, carbon (C) which has a high resistance with respect to sputtering, for instance, can be used. Here, when Sn is used as the target, for instance, as the material of the coating layers 21b and 22b, it is preferable to use titanium (Ti) which has a high wettability with respect to liquid Sn and a comparatively high resistance with respect to sputtering. This is because, by using porous Ti as a coating material, liquid Sn adhered to the surfaces of the coating layers 21b and 22b infiltrate into porus of the porous material. As a result, hardly any Sn will exist on the surface of the porous Ti, and thereby, the surfaces of the coating layers 21b and 22b coating the magnetic cores 21a and 22a may become hard to be sputtered even if Sn ion collide with the coating films 21b and 22b. The other structures and effects are the same as in the above-described embodiments, and redundant explanations will be omitted herein. Furthermore, in this particular embodiment, the case where the magnetic cores 21a and 22a are coated with the coating layers 21b and 22b has been described as an example. However, the present invention is not limited to such arrangement, while it is possible to coat members such as the ion collection (receiving) cylinders 23 and 24, and so on, having possibilities of being sputtered by ion debris, with the coating films 21b and 22b. Next, a seventh embodiment of the present invention will be described in detail. FIG. 16 is a horizontal cross-sectional diagram showing an outline structure of an extreme ultraviolet light source apparatus according to the seventh embodiment of the present invention. FIG. 17 is a vertical cross-sectional diagram showing an outline structure of the extreme ultraviolet light source apparatus according to the seventh embodiment. FIG. 18 is a diagram showing an example of a far field pattern formed on B-B surface in FIG. 16. As shown in FIGS. 16 to 18, the extreme ultraviolet light source apparatus according to the present embodiment has the same structure as the extreme ultraviolet light source apparatus according to the above-described fifth embodiment shown in FIGS. 12 to 14. However, in the extreme ultraviolet light source apparatus according to this particular embodiment, the magnetic cores 21a and 22a are omitted, while at least parts of the electromagnetic coils 21 and 22 are alternatively located in the obscuration region E (in particular, in the obscuration regions EDa and EDb). Even according to such structure, as in the case of the above-described fifth embodiment, it is possible to closely locate the electromagnetic coils 21 and 22 as being origins of the magnetic field in such a way that the electromagnetic coils 21 and 22 face closely to each other while sandwiching the plasma luminescence point P1 in between without shutting off the EUV light Eb to be used for exposure in the EUV exposure apparatus 11. Thereby, it is possible to locally form a strong magnetic field around the plasma luminescence point P1. As a result, it is possible to steadily trap ion debris generated around the plasma luminescence point P1 in the convergence region E2 and effectively lead the ion debris toward desired directions (directions toward the ion collection (receiving) cylinders 23 and 24). Furthermore, the ion collection (receiving) cylinders 23 and 24 are supporting the magnetic coils 21 and 22 by being extended inside the bores of the magnetic coils 21 and 22 from the wall surface of the vacuum chamber 1, respectively, and being fixed to the magnetic coils 21 and 22 inside the bores, respectively. Thereby, as in the case of the above-described fifth embodiment, it is possible to have the ion collection (receiving) cylinders 23 and 24 steadily collect (receive) ion debris drifting along the magnetic field passing through the bore of each of the electromagnetic coils 21 and 22. In this arrangement, as in the case of the above-described fifth embodiment, by locating at least parts of the ion collection (receiving) cylinders 23 and 24 inside the obscuration regions EDa and EDb in the EUV area Eb, it is possible to shorten a distance between the plasma luminescence point P1 and the ion collection (receiving) cylinders 23 and 24, respectively, and thus, it is possible to make a trapping solid angle with respect to an existence area of ion debris (the convergence region E2) larger, whereby it is possible to have the ion collection (receiving) cylinders 23 and 24 steadily collect (receive) ion debris generated around the plasma luminescence point P1. The other structures and effects are the same as the above-described embodiments, and redundant explanations will be omitted herein. Furthermore, as in the above-described sixth embodiment, in this particular embodiment also, it is preferable to have the electromagnetic coils 21 and 22 coated with the coating layers 21b and 22b. Thereby, it is possible to prevent conductive materials such as capper (Cu) constructing the electromagnetic coils 21 and 22 from being sputtered. In the above-described first to seventh embodiments, neutral particles, and so forth, in the convergence region E2 are not ionized. However, it is possible to facilitate trapping of debris by further mounting an ionization means such as an x-irradiator, an electron-irradiator, an ultraviolet irradiator, a microwave irradiator, an EUV light irradiator, or the like. Moreover, in each embodiment described above, the electromagnetic coils 21 and 22 are used as magnets that generate the magnetic field for trapping ion debris. However, the present invention is not limited to such arrangement, while permanent magnets can be used instead. According to each embodiment described above, due to having the ion debris removal unit mounted in the obscuration region including the plasma luminescence point where plasma is generated, it is possible to achieve such extreme ultraviolet light source apparatus that can output EUV light with desired intensity while being able to have debris being influential to optical element, and so on, collected (received) in the vacuum chamber using a simple structure. In addition, the above-mentioned embodiments and the alternate examples can be arbitrarily combined with one another. Additional advantages and modifications will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept of the invention as defined by the appended claims and their equivalents. |
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description | The present invention relates to an apparatus for passively cooling and retaining molten core material, and more particularly, to an apparatus for passively cooling and retaining molten core material, which can passively cool, solidify and retain ultrahigh-temperature radioactive molten core material, which is discharged from a damaged nuclear reactor vessel during a severe accident in a nuclear plant, in a reactor cavity so as to prevent the discharged molten core material from damaging a neighboring structure, threatening the safety of workers, threatening the integrity of the reactor containment, or contaminating water sources or soil. Generally, a nuclear plant includes hundreds of systems having separate functions. These systems are roughly classified into a nuclear steam supply system (NSSS) having a nuclear reactor as its main element, a turbine/generator system having a generator driven by supplied steam and other auxiliary equipments. Here, the nuclear reactor controls large amount of energy instantaneously emitted during a nuclear reaction to be slowly released so that nuclear energy can be utilized for generating electricity. However, when molten core material discharged from a damaged reactor vessel during a severe accident in the nuclear plant is not effectively retained and cooled, the discharged molten core material may ablate the concrete structure of a reactor cavity, contaminate neighboring soil, and threaten the structural integrity of a containment building. Accordingly, apparatuses and methods for retaining and suitably cooling the discharged molten core material have been constantly studied. FIG. 1 illustrates a conventional apparatus for retaining and cooling discharged molten core material, which is disclosed in a Korean Patent No. 0265320. In the conventional apparatus shown in FIG. 1, a multi-layered cell-type cooling water retention tank 110 is installed below a reactor vessel 100 and filled with cooling water 120 in order to sequentially cool ultrahigh-temperature molten core material discharged from the reactor vessel 100. However, the so-constructed conventional apparatus cannot be installed in a containment building for existing nuclear reactors because the containment building must be greatly altered in its compartment structure so as to install the apparatus therein. Also, the apparatus has a problem in that the ultrahigh-temperature molten core material melts the cooling water retention tank 110 and then reacts directly with the cooling water in the cooling water retention tank 110 thereby to generate steam violently, whereby a steam explosion may undesirably occur. As another conventional apparatus for retaining and cooling the discharged molten core material, FIG. 2 illustrates a core catcher having elongated molten core material catch parts, disclosed in Japanese Patent Publication No. 1996-43575. As shown in FIG. 2, the conventional apparatus has a plurality of molten core material catcher parts 130 installed below a reactor vessel 100. In the apparatus, molten core material discharged from the reactor vessel 100 during a severe accident in a nuclear plant is retained in the molten core material catcher parts 130, and the retained molten core material is cooled by cooling water 120 filled around the molten core material catcher parts 130. However, the above-described apparatus cools the molten core material of high temperature by bringing the molten core material into indirect contact with the cooling water. Accordingly, heat transfer efficiency between the molten core material and the cooling water is low and thus a very large cooling water tank is required. Also, there has been proposed an apparatus for actively cooling the molten core material by using a pump. However, the pump may not be operated when a severe accident occurs in a nuclear plant, whereby the apparatus has low reliability. Further, the cooling ability of the pump is degraded as time goes on, whereby the decay heat of the molten core material cannot be effectively removed. Accordingly, there is required an improved apparatus for passively cooling and retaining the molten core material, which can solve the above-described problems of the related art. One object of the present invention is to provide an improved apparatus for passively cooling and retaining molten core material, which can greatly reduce the possibility of steam explosion during the direct interaction between ultrahigh-temperature molten core material and cooling water, can effectively remove the decay heat of the molten core material, and can be assembled in a containment building for both existing nuclear reactors and newly constructed nuclear reactors. In order to achieve the above objects, the present invention provides an apparatus for passively cooling and retaining molten core material from a reactor, the apparatus comprising: a molten core material retention tank installed below a reactor vessel to retain molten core material discharged from the reactor vessel; a compressed gas tank having an outlet valve at an outlet thereof and supplying high-pressure inert gas; a cooling water storage tank being installed higher than the molten core material retention tank, having an outlet valve at an outlet thereof, and supplying cooling water, and a means for mixing inert gas supplied from the compressed gas tank with cooling water supplied from the cooling water storage tank and supplying the cooling water/inert gas mixture to the molten core material retention tank. The molten core material retention tank may include: an outer retention vessel having at least one coolant hole formed in a side or bottom thereof and connected to the mixing means; a porous protection vessel formed of refractory material at an inside of the outer retention vessel; and a gravel layer formed between the outer retention vessel and the porous protection vessel, and filled with refractory gravels. The gravels may be filled in the gravel layer to distribute and support the load of molten core material retained in the porous protection vessel. The mixing means may include pipes connected and extended respectively from the compressed gas tank and the cooling water storage tank. The porous protection vessel may be made by sintering refractory gravel or powder and on the inner surface of protection vessel, a layer of sacrificial and water tight material is cemented, and the outer retention vessel may have a screen layer formed on an inner surface thereof. The cooling water storage tank may have a check valve installed at the outlet thereof so as to prevent the back flow of high-pressure gas. The apparatus may further include an intermediate storage tank. Accordingly, steam generated by the reaction between the molten core material and the cooling water is condensed into water on the wall of the reactor containment and the condensed water is resupplied through the intermediate storage tank to the cooling water storage tank. The intermediate storage tank may have a filter installed in an upper side thereof to filter the condensed water through the filter, whereby the filtered water is resupplied to the cooling water storage tank. Hereinafter, preferred embodiments of the present invention will be described in detail with reference to the attached drawings. FIG. 3 is a schematic view of an apparatus for passively cooling and retaining molten core material according to the present invention. Referring to FIG. 3, the inventive passive cooling and retaining apparatus cools and retains molten core material from a reactor vessel, and includes a molten core material retention tank 20 installed below a reactor vessel 10 so as to retain molten core material discharged from the reactor vessel 10. The molten core material retention tank 20 is configured not only to endure high-temperature molten core material, but also to uniformly provide cooling water for uniformly cooling the high-temperature molten core material. A structure of the molten core material retention tank 20 will be described in detail later with reference to FIGS. 4 through 6. As shown in FIG. 3, the inventive passive cooling and retaining apparatus further includes a cooling water storage tank 40 installed in a containment building 1, and a compressed gas tank 30 for storing a high-pressure inert gas. The cooling water storage tank 40 and the compressed gas tank 30 respectively have outlet valves 41 and 31 at their outlets. Here, the cooling water storage tank 40 is preferably installed higher than the molten core material retention tank 20 so that the cooling water thereof can passively flow down into the molten core material retention tank 20 by gravity. Also, a check valve 41a is preferably installed at the outlet of the cooling water storage tank 40 so as to prevent the backflow of the high-pressure inert gas. Inert gas flowing down from the compressed gas tank 30 is mixed with cooling water flowing from the cooling water storage tank 40 by a mixer 50, and then supplied to the molten core material retention tank 20. Here, the mixer 50 is preferably of pipes connected and extended respectively from the tanks 30 and 40. Accordingly, the mixer 50 can simply mix the inert gas having flowed down from the tank 30 with the cooling water having flowed down from the tank 40, and the resulting mixture of the cooling water and the inert gas is supplied to both the inner bottom and side surfaces of the molten core material retention tank 20, whereby molten core material in the molten core material retention tank 20, whereby molten core material in the molten core material retention tank 20 can be more effectively cooled due to the large contact area between the molten core material and the supplied cooling water/inert gas mixture. As shown in FIG. 3, the passive cooling and retaining apparatus of the invention may further include an intermediate storage tank 60. Then, when steam generated by the reaction between the molten core material and the cooling water is condensed on the inner surface of the containment building 1, the condensed water can flow along the inner surface of the containment building 1 and then down via the intermediate storage tank 60 into the intermediate storage tank 60. In addition, the intermediate storage tank 60 preferably has a filter 61 installed in an upper side thereof to filter the condensed water. The condensed and filtered water is then recirculated via the intermediate storage tank 60 into the cooling water storage tank 40. In this case, the intermediate storage tank 60 is connected with the cooling water storage tank 40 via a pipe 62, by which the condensed water can be resupplied via the intermediate storage tank 60 into the cooling water storage tank 40. In this manner, the direct contact between the molten core material and the cooling water makes it possible to effectively remove the heat from the molten core material. Also, the inert gas supplied along with the cooling water can prevent steam explosion that may be generated by a violent interaction between the ultrahigh-temperature molten core material and the cooling water, thereby making it possible to stably cool the molten core material. Further, the invention is so .designed that the steam produced through the reaction between the molten core material and the cooling water can be condensed and the condensed water can be recycled through the intermediate storage tank, by which the decay heat can be efficiently removed from the molten core material. Furthermore, the inventive passive cooling device supplies the cooling Water and the inert gas and recycles the condensed water by gravity, thereby making it possible to cool the Molten core material more reliably than the conventional active cooling device using a pump. FIGS. 4 through 6 illustrate the molten core material retention tank in detail, which will be described as follows. Referring to FIG. 4, the molten core material retention tank 20 is connected through a coolant hole 21a with the mixer 50 of pipes, and is simultaneously supplied with the inert gas and the cooling water through the coolant hole 21a from the mixer 50, thereby cooling the high-temperature molten core material retained therein. Here, the molten core material retention tank 20 can more effectively cool the high-temperature molten core material due to the large contact area between the molten core material and the cooling water because the cooling water/inert gas mixture is supplied to both the inner bottom and side surfaces of the molten core material retention tank 20. Describing the structure of the invention in more detail referring to FIGS. 4 and 5, the molten core material retention tank 20 includes an outer retention vessel 21, an inner porous protection vessel 23, and a gravel layer 22 formed between the vessels 21 and 23. The outer retention vessel 21 includes at least one coolant hole 21a formed in a side or bottom surface thereof The coolant hole 21a in the bottom surface of the outer retention vessel 21 is depicted by an imaginary line in FIG. 5 to generally indicate that the bottom surface may include the coolant hole 21a rather than indicating a specific location for the coolant hole 21a in the bottom surface. The coolant hole 21a is connected to the mixer 50, whereby the cooling water/inert gas mixture is supplied through the coolant hole 21a to the molten core material retention tank 20. The inner porous protection vessel 23 inside the outer retention vessel 21 is preferably made of sintered refractory gravels or powder so as to endure the high-temperature molten core material. Accordingly, the cooling water/inert gas mixture can smoothly flow into the molten core material retention tank 20. The inner porous protection vessel 23 can be of several separate cylinder walls that can be assembled in a compartment structure. Also, the inner porous protection vessel 23 is made by firing refractory gravel or powder like a ceramic structure, and thus can be easily adjusted according to the shapes of the compartment. On the surface of the protection vessel, a layer of sacrificial material is cemented thick enough to provide the structural integrity and water tightness of the protection vessel 23. The sacrificial material has advantages of diluting the molten core material, having flexibility in responding to the various pattern of melt discharge from the reactor vessel. Also, as it provides water tightness of the protection vessel, inadvertent opening the valves 31, 41, and 41a during the power operation would not result in a flooding of the reactor cavity. So, any mis-operation of this invention would not have interference during the power operation of the nuclear reactor. So, it supports fully passive nature of the invention. The porous structure of the protection vessel 23 enables the cooling water/inert gas mixture to be uniformly injected through the pores thereof into the molten core material retention tank 20, and is highly endurable against the ultrahigh-temperature molten core material, thereby potentially preventing the direct contact between the ultrahigh-temperature molten core. material and a neighboring structure. Also, due to its structure, the molten core material retention tank 20 can be removed or replaced by a new one after a severe accident. Regarding the gravel layer 22, it is formed by filling refractory gravels 22a between the vessels 21 and 23, and thus can be easily formed to conform to the shape of the compartment due to the flowable nature of the refractory gravels 22a. The gravel layer 22 enables the cooling water/inert gas mixture to be uniformly injected into the molten core material retention tank 20. Also, the gravel layer 22 can store therein a predetermined amount of cooling water, thereby making it possible to more effectively cool the high-temperature molten core material. The gravel layer 22 is filled with the gravels 22a for supporting the weight of the molten core material retained in the porous protection vessel 23. That is, the gravels 22a of the gravel layer 22 themselves can distribute and support the weight of the molten core material. Accordingly, the present invention can support the load of the molten core material without having to use a separate supporting unit. Also, the supply amount of the cooling water/inert gas mixture can be suitably adjusted by changing the arrangement of the coolant hole 21a and the gravels 22a. That is, the gravel 22a size of the gravel layer 22 and the pore size of the porous protection vessel 23 can be adjusted suitably according to the size of a reactor cavity and the total amount of the molten core material, whereby the molten core material retention tank 20 can be optimally supplied with the cooling water/inert gas mixture. FIG. 6 is a perspective view of the outer retention vessel 21 of the molten core material retention tank 20. Referring to FIG. 6, the outer retention vessel 21 preferably has a screen layer 21b formed on an inner surface thereof to border the gravel layer 22. The screen layer 21b is designed to prevent the gravels 22a from emerging through the coolant hole 21a, and supplying the cooling water/inert gas mixture more uniformly. An operation of the inventive passive cooling and retaining apparatus will now be described in detail with reference to FIGS. 3 through 6. Referring to FIGS. 3 through 6, when a severe accident takes place in a nuclear plant, discharging ultrahigh-temperature radioactive molten core material from a damaged nuclear reactor into a reactor cavity, the discharged molten core material is retained in the molten core material retention tank 20 installed below the reactor vessel 10. When the outlet valves 31 and 41 are opened, the inert gas and the cooling water respectively flow down from the compressed gas tank 30 and the cooling water storage tank 40 by pressure and gravity and are mixed in the mixer 50. The cooling water/inert gas mixture is then supplied to the molten core material retention tank 20, when the molten material discharged from the reactor vessel is accumulated enough to cause a failure of a sacrificial water tight layer on the inner surface of the protection vessel 23. At this time, the cooling water storage tank 40 is preferably installed higher than the molten core material retention tank 20 so that the cooling water thereof can be smoothly supplied to the molten core material retention tank 20 by gravity. Also, the check valve 41a is preferably installed at the outlet of the cooling water storage tank 40 so as to prevent the back flow of the high-pressure inert gas. Here, the molten core material retention tank 20 includes the outer retention vessel 21, the inner porous protection vessel 23 and the gravel layer 22 formed between the vessels 21 and 23 as shown in FIGS. 4 and 5. The outer retention vessel 21 includes at least one coolant hole 21a formed in a side or bottom surface thereof. The coolant hole 21a is connected to the mixer 50, whereby the cooling water/inert gas mixture is supplied through the coolant hole 21a to the molten core material retention tank 20. The porous protection vessel 23 and the gravel layer 22 are preferably made of refractory material so as to endure the high-temperature molten core material. The porous structures of the protection vessel 23 and the gravel layer 22 enable the uniform supply of the cooling water/inert gas mixture. After the molten core material is primarily cooled in the molten core material retention tank 20 as stated above, steam generated by the reaction between the molten core material and the cooling water is emitted into the atmosphere over the containment building 1 and is then condensed on an upper outer surface of the containment building 1. The condensed water is resupplied through the intermediate storage tank 60 and the pipe 62 to the cooling water storage tank 40. At this time, the intermediate storage tank 60 preferably has the filter 61 installed on an upper side thereof, and thus the condensed water is filtered by the filter 61 before flowing into the intermediate storage tank 60. Also, a feedwater line 70 for suitably supplying feedwater is connected to the cooling water storage tank 40. The cooling water/inert gas mixture is supplied to both the bottom and side surfaces of the molten core material by opening the outlet valves 31 and 41 before the solidification of high-temperature molten core material, whereby the molten core material is primarily cooled so that a crack-type flow channel can be formed in a solidified molten core material. Thereafter, cooling water is supplied to the crack-type flow channel of the solidified molten core material, whereby the decay heat of the solidified molten core material is removed. As stated above, the present invention can passively supply the cooling water/inert gas mixture during the primary cooling process thereby to greatly reduce the threat of a steam explosion, which may be caused by the rapid reaction between the high-temperature molten core material and the cooling water. Also, the present invention makes it possible to effectively remove the decay heat of the molten core material by passively recycling the condensed water. Further, the present invention makes it possible to separately make and assembly the molten core material retention tank suitably according to the size of the reactor cavity, and to install the molten core material retention tank in the reactor cavity even for an existing nuclear reactor. Because the sacrificial and water tight layer is cemented on the protection vessel 23, inadvertent opening the valves 31, 41, and 41a during the power operation would not result in a flooding of the reactor cavity. So, any mis-operation of this invention would not interfere with the normal power operation of the nuclear reactor. Accordingly, this feature fully supports the passive nature of the invention. Although the preferred embodiments of the present invention have been disclosed for illustrative purpose, those skilled in the art will appreciate that various modifications, additions and substitutions can be made without departing from the scope and spirit of the invention as defined in the accompanying claims. The inventive passive cooling and retaining apparatus can prevent steam explosion by supplying the cooling water/inert mixture to both the bottom and side surfaces of the ultrahigh-temperature molten core material, thereby making it possible to greatly enhance the reliability of the cooling process. Also, the present invention makes it possible to install the molten core material retention tank in the reactor cavity without changing the compartment structure of the containment building of the nuclear plant even for the existing nuclear reactor and it can be installed easily in the newly constructed nuclear reactor. Further, the molten core material retention tank, of this invention includes the outer retention vessel, the inner porous protection vessel, and the gravel layer formed between the vessels. Accordingly, the cooling water/inert gas mixture can be uniformly injected into the molten core material retention tank, a predetermined amount of cooling water can be stored in the gravel layer, and the gravel layer can distribute and support the weight of the molten core material. Furthermore, the present invention makes it possible to effectively remove the decay heat of the molten core material by passively recycling the cooling water through the intermediate storage tank. As the protection vessel with a layer of sacrificial material provides water tightness and structural integrity, any mis-operation of this invention would not interfere with the normal power operation of the nuclear reactor. |
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abstract | A lithographic apparatus includes a source configured to generate a radiation beam comprising desired radiation and undesired radiation using a plasma, an illumination system configured to condition the radiation beam and to receive hydrogen gas during operation of the lithographic apparatus, and a support structure constructed to hold a patterning device. The patterning device is capable of imparting the radiation beam with a pattern in its cross-section to form a patterned radiation beam. A substrate table is constructed to hold a substrate, and a projection system is configured to project the patterned radiation beam onto a target portion of the substrate. The lithographic apparatus is configured such that the radiation beam on entering the projection system includes at least 50% of the undesired radiation that is generated by the plasma and includes wavelengths of radiation that interact with the hydrogen gas to generate hydrogen radicals. |
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abstract | A charged particle microscope corrects distortion in an image caused by effects of drift in the sampling stage by measuring the correction reference image in a shorter time than the observation image, making corrections by comparing the shape of the observation image with the shape of the correction reference image, and reducing distortion in the observation images. The reference image for distortion correction is measured at the same position and magnification as when acquiring images for observation. In order to reduce effects from drift, the reference image is at this time measured within a shorter time than the essential observation image. The shape of the observation image is corrected by comparing the shapes of the reference image and observation image, and correcting the shape of the observation image to match the reference image. |
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description | This application is a continuation of U.S. patent application Ser. No. 13/830,461, filed Mar. 14, 2013, now allowed, and titled TARGET FOR LASER PRODUCED PLASMA EXTREME ULTRAVIOLET LIGHT SOURCE, which is incorporated herein by reference in its entirety. The disclosed subject matter relates to a target for a laser produced plasma extreme ultraviolet light source. Extreme ultraviolet (“EUV”) light, for example, electromagnetic radiation having wavelengths of around 50 nm or less (also sometimes referred to as soft x-rays), and including light at a wavelength of about 13 nm, can be used in photolithography processes to produce extremely small features in substrates, for example, silicon wafers. Methods to produce EUV light include, but are not necessarily limited to, converting a material that has an element, for example, xenon, lithium, or tin, with an emission line in the EUV range in a plasma state. In one such method, often termed laser produced plasma (“LPP”), the required plasma can be produced by irradiating a target material, for example, in the form of a droplet, plate, tape, stream, or cluster of material, with an amplified light beam that can be referred to as a drive laser. For this process, the plasma is typically produced in a sealed vessel, for example, a vacuum chamber, and monitored using various types of metrology equipment. In one general aspect, a method of generating EUV light includes directing a first pulse of radiation toward a target material droplet to form a modified droplet, the first pulse of radiation having an energy sufficient to alter a shape of the target material droplet; directing a second pulse of radiation toward the modified droplet to form an absorption material, the second pulse of radiation having an energy sufficient to change a property of the modified droplet, the property being related to absorption of radiation; and directing an amplified light beam toward the absorption material, the amplified light beam having an energy sufficient to convert at least a portion of the absorption material into extreme ultraviolet (EUV) light. Implementations can include one or more of the following features. The modified droplet can include a continuous segment of the target material that has a width extending along a first direction, and a thickness extending along a second direction that is different from the first direction. The second direction is in the direction of propagation of the second pulse of radiation, and the width is greater than the thickness. A plane that includes the first direction can be angled relative to the direction of propagation of the second pulse of radiation. The absorption material can include a continuous segment of the target material. The property of the modified droplet can be one or more of an electron density and an ion density, and the absorption material can include plasma adjacent to a surface of a continuous segment of the target material. The property of the modified droplet can be a surface area. The absorption material can include multiple pieces of the target material, the multiple pieces having a collective surface area that is larger than the modified droplet. The first pulse of radiation can be a pulse of light having a wavelength of 10 μm, a pulse duration of 40 ns, and an energy of 20 mJ, and the second pulse of radiation can be a pulse of light wavelength of 1 μm, a pulse duration of 10 ns, and an energy of 5 mJ. The first pulse of radiation can be a pulse of light having a wavelength of 10 μm, a pulse duration of 20-70 ns, and an energy of 15-60 mJ, and the second pulse of radiation can be a pulse of light wavelength of 1-10 μm, a pulse duration of 10 ns, and an energy of 1-10 mJ. The first pulse of radiation can be a pulse of light having a wavelength of 1-10 μm, a pulse duration of 40 ns, and an energy of 20 mJ, and the second pulse of radiation can be a pulse of light having a wavelength of 1 μm, a pulse duration of 10 ns, and an energy of 1 mJ. The first pulse of radiation and the second pulse of radiation can be pulses of light having a duration of 1 ns or greater. The second pulse of radiation can be a pulse of light having a duration of 1 ns to 100 ns. The second pulse of radiation can be directed toward the modified droplet 1-3 μs after the first pulse of radiation is directed toward the target material droplet. The first pulse of radiation can be a pulse of light having a duration of at least 1 ns, and the second pulse of radiation can be a pulse of light having a duration of at least 1 ns. In some implementations, at least 2% of the amplified light beam can be converted to EUV radiation. The amplified light beam can be a pulse of light, and a subsequent pulse of light can be directed toward a second absorption material no more than 25 μs after the amplified light beam is directed toward the absorption material. The second absorption material is formed after the absorption material and is formed from a second target material droplet. The first pulse of radiation can be a pulse of radiation having a duration of 300 ps or less. The first pulse of radiation can be a pulse of radiation having a duration of 100 ps-300 ps. The modified droplet can be a hemisphere shaped volume of particles of target material. In another general aspect, an extreme ultraviolet light source includes a source that produces an amplified light beam, a first pulse of radiation, and a second pulse of radiation; a target material delivery system; a vacuum chamber coupled to the target material delivery system; and a steering system configured to steer and focus the amplified light beam, the first pulse of radiation, and the second pulse of radiation toward a target location that receives target material from the target material delivery system in the vacuum chamber. The first pulse of radiation has an energy sufficient to alter a shape of the target material droplet to create a modified droplet, the second pulse of radiation has an energy sufficient to change a property of the modified droplet that is related to absorption of radiation, and the amplified light beam is sufficient to convert at least a portion of the absorption material into extreme ultraviolet (EUV) light. Implementations can include one or more of the following features. The source can include first, second, and third sources, with the first source generating the first pulse of radiation, the second source generating the second pulse of radiation, and the third source generating the amplified light beam. The source can include a first source that generates the amplified light beam and the first pulse of radiation and a second source that generates the second pulse of radiation. The first source can include a CO2 laser, and the amplified light beam and the first pulse of radiation can have different wavelengths. Implementations of any of the techniques described above may include a target for a laser produced plasma EUV light source, an EUV light source, a system for retrofitting an EUV light source, a method, a process, a device, executable instructions stored on a computer readable medium, or an apparatus. The details of one or more implementations are set forth in the accompanying drawings and the description below. Other features will be apparent from the description and drawings, and from the claims. Techniques for producing a target for use in a laser produced plasma (LPP) extreme ultraviolet (EUV) light source are disclosed. The target is produced by irradiating a target material with two pulses of light in succession. The first pulse generates an intermediate target and the second pulse interacts with the intermediate target to produce the target. The target is then irradiated with an amplified light beam having energy that is sufficient to convert target material in the target to a plasma that emits EUV light. In some implementations, each of the two pulses of light has a temporal duration or pulse width of at least 1 nanosecond (ns). Referring to FIGS. 1A and 1B, an exemplary waveform 5 transforms a target material 50 into a target 55. The target 55 includes target material that emits EUV light 57 when converted to plasma. The target material 50 can be a target mixture that includes a target substance and impurities such as non-target particles. The target substance is the substance that is converted to a plasma state that has an emission line in the EUV range. The target substance can be, for example, a droplet of liquid or molten metal, a portion of a liquid stream, solid particles or clusters, solid particles contained within liquid droplets, a foam of target material, or solid particles contained within a portion of a liquid stream. The target substance, can be, for example, water, tin, lithium, xenon, or any material that, when converted to a plasma state, has an emission line in the EUV range. For example, the target substance can be the element tin, which can be used as pure tin (Sn); as a tin compound, for example, SnBr4, SnBr2, SnH4; as a tin alloy, for example, tin-gallium alloys, tin-indium alloys, tin-indium-gallium alloys, or any combination of these alloys. Moreover, in the situation in which there are no impurities, the target material includes only the target substance. The discussion below provides an example in which the target material 50 is a target material droplet made of molten metal. The target material 50 is referred to as the target material droplet 50. However, the target material 50 can take other forms. FIG. 1A shows the target material droplet 50 physically transforming into an intermediate target 51 and then into the target 55 over a time period. The target material droplet 50 is transformed through interaction with the radiation delivered in time according to the waveform 5. FIG. 1B is a plot of the energy in the waveform 5 as a function of time over the time period of FIG. 1A. As compared to the target material droplet 50 and the intermediate target 51, the target 55 absorbs more of an amplified light beam 8 and converts a larger portion of the energy in the amplified light beam 8 to the EUV light 57. The waveform 5 is a representation of the energy that interacts with the target material droplet 50 and its modified forms over time. Although the waveform 5 is shown as a single waveform as a function of time, various portions of the waveform 5 can be produced by different sources. The waveform 5 includes a representation of a first pulse of radiation 6 (a first pre-pulse 6) and a representation of a second pulse of radiation 7 (a second pre-pulse 7). The first pre-pulse 6 and the second pre-pulse 7 can be any type of pulsed radiation that has sufficient energy to act on the target material droplet 50 and the intermediate target 51, respectively. Examples of pre-pulses are discussed with respect to FIGS. 3A-3D, 4, 5, 7 and 9. The first pre-pulse 6 occurs at a time t=t1 and has a pulse duration 12, and the second pre-pulse 7 occurs at a time t=t2 and has a pulse duration 14. The pulse duration can be represented by the full width at half maximum, the amount of time that the pulse has an intensity that is at least half of the maximum intensity of the pulse. However, other metrics can be used to determine the pulse duration. The times t1 and t2 are separated by a first delay time 11, with the second pre-pulse 7 occurring after the first pre-pulse 6. The waveform 5 also shows a representation of the amplified light beam 8. The amplified light beam 8 can be referred to as the main beam or the main pulse. The amplified light beam 8 has sufficient energy to convert target material in the target 55 to plasma that emits EUV light. The second pre-pulse 7 and the amplified light beam 8 are separated in time by a second delay time 13, with the amplified light beam 8 occurring after the second pre-pulse 7. Examples of an EUV light source that can produce and/or use the target 55 are shown in FIGS. 2A, 2B, and 3A-3E. Before discussing the EUV light sources, a discussion of the interactions of the pulses of light, including the first pre-pulse 6 and the second pre-pulse 7, with the target material droplet 50 and the intermediate target 51 is provided. When a laser pulse impinges (strikes) a target material droplet that is metallic, the leading edge of the pulse sees (interacts with) a surface that is a reflective metal. The target material droplet 50 reflects most of the energy in the leading edge of the pulse and absorbs little. The small amount that is absorbed heats the surface of the droplet, evaporating and ablating the surface. The target material that is evaporated from the surface of the droplet forms a cloud of electrons and ions close to the surface. As the pulse of radiation continues to impinge on the target material droplet, the electric field of the laser pulse can cause the electrons in the cloud to move. The moving electrons collide with nearby ions, heating the ions through the transfer of kinetic energy at a rate that is roughly proportional to the product of the densities of the electrons and the ions in the cloud. Through the combination of the moving electrons striking the ions and the heating of the ions, the cloud absorbs the pulse. As the cloud is exposed to the later parts of the laser pulse, the electrons in the cloud continue to move and collide with ions, and the ions in the cloud continue to heat. The electrons spread out and transfer heat to the surface of the target material droplet (or bulk material that underlies the cloud), further evaporating the surface of the target material droplet. The electron density in the cloud increases in the portion of the cloud that is closest to the surface of the target material droplet. The cloud can reach a point where the density of electrons increases such that the portions of the cloud reflect the laser pulse instead of absorbing it. The present technique for generating a target for an LPP EUV light source applies two pre-pulses to a target material droplet to physically transform the target material droplet into a target that more readily absorbs energy. The first pre-pulse 6 forms a geometric distribution of target material that becomes the intermediate target 51. The second pre-pulse 7 transforms the intermediate target 51 into the target 55. The first pre-pulse 6 and the second pre-pulse 7 are discussed in turn below. Referring also to FIG. 1C, the first pre-pulse 6 physically transforms the target material droplet into a geometric distribution 52 of target material. The geometric distribution 52 can be a material that is not ionized (a material that is not a plasma). The geometric distribution 52 can be, for example, a disk of liquid or molten metal, a continuous segment of target material that does not have voids or substantial gaps, a mist of micro- or nano-particles, or a cloud of atomic vapor. The geometric distribution 52 expands spatially during the first delay time 11 and becomes the intermediate target 51. The first pre-pulse 6 spreads the target material droplet 50 spatially. Spreading the target material droplet 50 can have two effects. First, the intermediate target 51 generated by the first pre-pulse 6 has a form that presents a larger area to an oncoming pulse of radiation (such as the pre-pulse 7). The intermediate target 51 has a cross-sectional diameter 54 that is larger than a beam diameter 57 of the pre-pulse 7 such that the intermediate target receives the entire pre-pulse 7. Additionally, the intermediate target 51 can have a thickness 58 that is thinner in a direction of propagation of the pre-pulse 7 than a thickness 59 of the target material droplet 50. The relative thinness of the intermediate target 51 allows the pre-pulse beam 7 to irradiate more of the target material that is in the intermediate target 51, including more of the target material that is not irradiated by the pre-pulse 7 when it initially reaches the intermediate target 51. Second, spreading the target material of the droplet 50 out spatially can minimize the occurrence of regions of excessively high material density during heating of the plasma by the strong pulse 8, which can block generated EUV light. If the plasma density is high throughout a region that is irradiated with a laser pulse, absorption of the laser pulse is limited to the portions of the region that receives the laser pulse first. Heat generated by this absorption may be too distant from the bulk target material to maintain the process of evaporating and heating of the target material surface long enough to utilize (evaporate) a meaningful amount of the bulk target material during the finite duration of the pulse 8. In instances where the region has a high electron density, the light pulse only penetrates a fraction of the way into the region before reaching a “critical surface” where the electron density is so high that the light pulse is reflected. The light pulse cannot travel into those portions of the region and little EUV light is generated from target material in those regions. The region of high plasma density can also block EUV light that is emitted from the portions of the region that do emit EUV light. Consequently, the total amount of EUV light that is emitted from the region is less than it would be if the region lacked the portions of high plasma density. As such, spreading the target material droplet 10 into the larger volume of the intermediate target 51 means that an incident light beam reaches more of the material in the intermediate target 51 before being reflected. This can increase the amount of EUV light subsequently produced. The waveform 5 also shows a representation of the second pre-pulse 7. The second pre-pulse 7 impinges on the intermediate target 51 and forms the target 55 before the amplified light beam 8 arrives. The target 55 can take many forms. For example, the target 55 can be a pre-plasma that is spatially near to a bulk target material. A pre-plasma is a plasma that is used to enhance absorption of incident light (such as the pre-pulse 7 or the amplified light beam). Although the pre-plasma can emit small amounts of EUV light in some instances, the EUV light that is emitted is not of the wavelength or amount that is emitted by the target 55. In other implementations, the target 55 is a volume of fragments or a mist of target material. An example of a waveform that includes a second pre-pulse that can form a pre-plasma is discussed below with respect to FIG. 5. An example of a waveform that includes a second pre-pulse that can form fragments of target material is discussed below with respect to FIG. 7. In yet other implementations, the target 55 is a pre-plasma formed close to a collection of particles of target material distributed throughout a hemisphere shaped volume. An example of such a target is discussed below with respect to FIG. 9. In some implementations, the pulse duration 12 of the first pre-pulse 6 and the pulse duration 14 of the second pre-pulse 7 are 1 ns or greater. Using two pre-pulses that are greater than 1 ns allows the target 55 to be produced using pulses of radiation that are generated without using a laser that generates picosecond (ps) or shorter pulses. Lasers that emit ns-duration pulses and have relatively high repetition rates (50 kHz-100 kHz) can be more readily available than those that emit ps-pulses. Use of higher-repetition rate ns-pulse generating lasers to generate the pre-pulses 6 and 7 allows an EUV light source that uses the target 55 to have a higher overall system repetition rate. FIG. 1D shows an exemplary plot of a waveform 60 over two continuous cycles of an EUV light source. The waveform 60 is two instances of the waveform 5 (FIG. 1A), with each cycle of the EUV light source applying an instance of the waveform 5 to two separate target material droplets (one per cycle) to emit EUV light once per cycle. In the example shown in FIG. 1D, EUV light emissions 61 and 62 occur after an instance of the waveform 5 is applied to a target material droplet. The emissions 61 and 62 are separated in time by a time 64 that is the inverse of the repetition rate of the EUV light source. The repetition rate of the EUV light source also can be considered as the minimum amount of time between two successive EUV light emissions. Because the time between the EUV light emissions 61 and 62 depends on how quickly instances of the waveform 5 can be generated, the repetition rate of the sources that generate the pre-pulses 6 and 7 at least partially determines the system repetition rate. When using two ns-duration pulses as the pre-pulses 6 and 7, the EUV light source's system repetition rate can be, for example, 40 kHz-100 kHz. Although the example of FIG. 1D shows continuous emission of EUV light, where EUV light is emitted at periodic intervals determined by the system repetition rate, the EUV light source can be operated in other modes depending on the needs of a lithography tool that receives the generated EUV light. For example, the EUV light source also can be operated or set to emit EUV light in bursts that are separated in time by an amount greater than the system repetition rate or at an irregular interval. The system repetition rate discussed with respect to FIG. 1A is provided as an example of a minimum amount of time between EUV light emissions. FIGS. 2A, 2B, and 3A-3C show exemplary LPP EUV light sources in which the target 55 can be used. Referring to FIG. 2A, an LPP EUV light source 100 is formed by irradiating a target mixture 114 at a target location 105 with an amplified light beam 110 that travels along a beam path toward the target mixture 114. The target location 105, which is also referred to as the irradiation site, is within an interior 107 of a vacuum chamber 130. When the amplified light beam 110 strikes the target mixture 114, a target material within the target mixture 114 is converted into a plasma state that has an element with an emission line in the EUV range. The created plasma has certain characteristics that depend on the composition of the target material within the target mixture 114. These characteristics can include the wavelength of the EUV light produced by the plasma and the type and amount of debris released from the plasma. The light source 100 also includes a target material delivery system 125 that delivers, controls, and directs the target mixture 114 in the form of liquid droplets, a liquid stream, solid particles or clusters, solid particles contained within liquid droplets or solid particles contained within a liquid stream. The target mixture 114 can also include impurities such as non-target particles. The target mixture 114 is delivered by the target material delivery system 125 into the interior 107 of the chamber 130 and to the target location 105. The light source 100 includes a drive laser system 115 that produces the amplified light beam 110 due to a population inversion within the gain medium or mediums of the laser system 115. The light source 100 includes a beam delivery system between the laser system 115 and the target location 105, the beam delivery system including a beam transport system 120 and a focus assembly 122. The beam transport system 120 receives the amplified light beam 110 from the laser system 115, and steers and modifies the amplified light beam 110 as needed and outputs the amplified light beam 110 to the focus assembly 122. The focus assembly 122 receives the amplified light beam 110 and focuses the beam 110 to the target location 105. In some implementations, the laser system 115 can include one or more optical amplifiers, lasers, and/or lamps for providing one or more main pulses and, in some cases, one or more pre-pulses. Each optical amplifier includes a gain medium capable of optically amplifying the desired wavelength at a high gain, an excitation source, and internal optics. The optical amplifier may or may not have laser mirrors or other feedback devices that form a laser cavity. Thus, the laser system 115 produces an amplified light beam 110 due to the population inversion in the gain media of the laser amplifiers even if there is no laser cavity. Moreover, the laser system 115 can produce an amplified light beam 110 that is a coherent laser beam if there is a laser cavity to provide enough feedback to the laser system 115. The term “amplified light beam” encompasses one or more of: light from the laser system 115 that is merely amplified but not necessarily a coherent laser oscillation and light from the laser system 115 that is amplified (externally or within a gain medium in the oscillator) and is also a coherent laser oscillation. The optical amplifiers in the laser system 115 can include as a gain medium a filling gas that includes CO2 and can amplify light at a wavelength of between about 9100 and about 11000 nm, and in particular, at about 10.6 μm, at a gain greater than or equal to 1000. In some examples, the optical amplifiers amplify light at a wavelength of 10.59 μm. Suitable amplifiers and lasers for use in the laser system 115 can include a pulsed laser device, for example, a pulsed, gas-discharge CO2 laser device producing radiation at about 9300 nm or about 10600 nm, for example, with DC or RF excitation, operating at relatively high power, for example, 10 kW or higher and high pulse repetition rate, for example, 50 kHz or more. The optical amplifiers in the laser system 115 can also include a cooling system such as water that can be used when operating the laser system 115 at higher powers. FIG. 2B shows a block diagram of an example drive laser system 180. The drive laser system 180 can be used as the drive laser system 115 in the source 100. The drive laser system 180 includes three power amplifiers 181, 182, and 183. Any or all of the power amplifiers 181, 182, and 183 can include internal optical elements (not shown). The power amplifiers 181, 182, and 183 each include a gain medium in which amplification occurs when pumped with an external electrical or optical source. Light 184 exits from the power amplifier 181 through an output window 185 and is reflected off a curved mirror 186. After reflection, the light 184 passes through a spatial filter 187, is reflected off of a curved mirror 188, and enters the power amplifier 182 through an input window 189. The light 184 is amplified in the power amplifier 182 and redirected out of the power amplifier 182 through an output window 190 as light 191. The light 191 is directed toward the amplifier 183 with fold mirrors 192 and enters the amplifier 183 through an input window 193. The amplifier 183 amplifies the light 191 and directs the light 191 out of the amplifier 183 through an output window 194 as an output beam 195. A fold mirror 196 directs the output beam 195 upwards (out of the page) and toward the beam transport system 120. The spatial filter 187 defines an aperture 197, which can be, for example, a circle through which the light 184 passes. The curved mirrors 186 and 188 can be, for example, off-axis parabola mirrors with focal lengths of about 1.7 m and 2.3 m, respectively. The spatial filter 187 can be positioned such that the aperture 197 coincides with a focal point of the drive laser system 180. The example of FIG. 2B shows three power amplifiers. However, more or fewer power amplifiers can be used. Referring again to FIG. 2A, the light source 100 includes a collector mirror 135 having an aperture 140 to allow the amplified light beam 110 to pass through and reach the target location 105. The collector mirror 135 can be, for example, an ellipsoidal mirror that has a primary focus at the target location 105 and a secondary focus at an intermediate location 145 (also called an intermediate focus) where the EUV light 106 can be output from the light source 100 and can be input to, for example, an integrated circuit beam positioning system tool (not shown). The light source 100 can also include an open-ended, hollow conical shroud 150 (for example, a gas cone) that tapers toward the target location 105 from the collector mirror 135 to reduce the amount of plasma-generated debris that enters the focus assembly 122 and/or the beam transport system 120 while allowing the amplified light beam 110 to reach the target location 105. For this purpose, a gas flow can be provided in the shroud that is directed toward the target location 105. The light source 100 can also include a master controller 155 that is connected to a droplet position detection feedback system 156, a laser control system 157, and a beam control system 158. The light source 100 can include one or more target or droplet imagers 160 that provide an output indicative of the position of a droplet, for example, relative to the target location 105 and provide this output to the droplet position detection feedback system 156, which can, for example, compute a droplet position and trajectory from which a droplet position error can be computed either on a droplet by droplet basis or on average. The droplet position detection feedback system 156 thus provides the droplet position error as an input to the master controller 155. The master controller 155 can therefore provide a laser position, direction, and timing correction signal, for example, to the laser control system 157 that can be used, for example, to control the laser timing circuit and/or to the beam control system 158 to control an amplified light beam position and shaping of the beam transport system 120 to change the location and/or focal power of the beam focal spot within the chamber 130. The target material delivery system 125 includes a target material delivery control system 126 that is operable in response to a signal from the master controller 155, for example, to modify the release point of the droplets as released by a target material supply apparatus 127 to correct for errors in the droplets arriving at the desired target location 105. Additionally, the light source 100 can include a light source detector 165 that measures one or more EUV light parameters, including but not limited to, pulse energy, energy distribution as a function of wavelength, energy within a particular band of wavelengths, energy outside of a particular band of wavelengths, and angular distribution of EUV intensity and/or average power. The light source detector 165 generates a feedback signal for use by the master controller 155. The feedback signal can be, for example, indicative of the errors in parameters such as the timing and focus of the laser pulses to properly intercept the droplets in the right place and time for effective and efficient EUV light production. The light source 100 can also include a guide laser 175 that can be used to align various sections of the light source 100 or to assist in steering the amplified light beam 110 to the target location 105. In connection with the guide laser 175, the light source 100 includes a metrology system 124 that is placed within the focus assembly 122 to sample a portion of light from the guide laser 175 and the amplified light beam 110. In other implementations, the metrology system 124 is placed within the beam transport system 120. The metrology system 124 can include an optical element that samples or re-directs a subset of the light, such optical element being made out of any material that can withstand the powers of the guide laser beam and the amplified light beam 110. A beam analysis system is formed from the metrology system 124 and the master controller 155 since the master controller 155 analyzes the sampled light from the guide laser 175 and uses this information to adjust components within the focus assembly 122 through the beam control system 158. Thus, in summary, the light source 100 produces an amplified light beam 110 that is directed along the beam path to irradiate the target mixture 114 at the target location 105 to convert the target material within the mixture 114 into plasma that emits light in the EUV range. The amplified light beam 110 operates at a particular wavelength (that is also referred to as a source wavelength) that is determined based on the design and properties of the laser system 115. Additionally, the amplified light beam 110 can be a laser beam when the target material provides enough feedback back into the laser system 115 to produce coherent laser light or if the drive laser system 115 includes suitable optical feedback to form a laser cavity. Referring to FIG. 3A, a top plan view of an exemplary optical imaging system 300 is shown. The optical imaging system 300 includes an LPP EUV light source 305 that provides EUV light to a lithography tool 310. The light source 305 can be similar to, and/or include some or all of the components of, the light source 100 of FIGS. 2A and 2B. As discussed below, the target 55 can be used in the light source 305 to increase the amount of light emitted by the light source 305. The light source 305 includes a drive laser system 315, an optical element 322, a pre-pulse source 324, a focusing assembly 326, a vacuum chamber 340, and a EUV collecting optic 346. The EUV collecting optic 346 directs the EUV light emitted by the target 55 to the lithography tool 310. The EUV collecting optic 346 can be the collector mirror 135 of FIG. 2A. The drive laser system 315 produces an amplified light beam 316. The amplified light beam 316 can be similar to the amplified light beam 18 of FIGS. 1A-1C and can be referred to as a main pulse or a main beam. The amplified light beam 316 has an energy sufficient to convert target material in the target 55 into plasma that emits EUV light. The pre-pulse source 324 emits pulses of radiation 317 and 318. The pulses of radiation 317 and 318 can be similar to the first pre-pulse 6 and the second pre-pulse 7 of FIG. 1B. The pre-pulse source 324 can be, for example, a Q-switched Nd:YAG laser that operates at a 50 kHz repetition rate, and the pulses of radiation 317 and 318 can be pulses from the Nd:YAG laser that have a wavelength of 1.06 μm. The repetition rate of the pre-pulse source 324 indicates how often the pre-pulse source 324 produces a pulse of radiation. For the example where the pre-pulse source 324 has a 50 kHz repetition rate, a pulse of radiation 317 is emitted from the source 324 every 20 microseconds (μs). Other sources can be used as the pre-pulse source 324. For example, the pre-pulse source 324 can be any rare-earth-doped solid state laser other that an Nd:YAG, such as an erbium-doped fiber (Er:glass) laser. The pre-pulse source 324 can be any other radiation or light source that produces light pulses that have an energy and wavelength used for the first pre-pulse 6 and the second pre-pulse 7. The optical element 322 directs the amplified light beam 316 and the pulses of radiation 317 and 318 from the pre-pulse source 324 to the chamber 340. The optical element 322 is any element that can direct the amplified light beam 316 and the pulses of radiation 317 and 318 along similar paths and deliver the amplified light beam 316 and the pulses of radiation 317 and 318 to the chamber 340. In the example shown in FIG. 3A, the optical element 322 is a dichroic beamsplitter that receives the amplified light beam 316 and reflects it toward the chamber 340. The optical element 322 receives the pulses of radiation 317 and 318 and transmits the pulses toward the optical chamber 340. The dichroic beamsplitter has a coating that reflects the wavelength(s)s of the amplified light beam 316 and transmits the wavelength(s) of the pulses of radiation 317 and 318. The dichroic beamsplitter can be made of, for example, diamond. In other implementations, the optical element 322 is a mirror that defines an aperture (not shown). In this implementation, the amplified light beam 316 is reflected from the mirror surface and directed toward the chamber 340, and the pulses of radiation pass through the aperture and propagate toward the chamber 340. In still other implementations, a wedge-shaped optic (for example, a prism) can be used to separate the main pulse 316, the pre-pulse 317, and the pre-pulse 318 into different angles, according to their wavelengths. The wedge-shaped optic can be used in addition to the optical element 322, or it can be used as the optical element 322. The wedge-shaped optic can be positioned just upstream (in the “−z” direction) of the focusing assembly 326. Additionally, the pulses 317 and 318 can be delivered to the chamber 340 in other ways. For example, the pulses 317 and 318 can travel through optical fibers that deliver the pulses 317 and 318 to the chamber 340 and/or the focusing assembly 326 without the use of the optical element 322 or other directing elements. In these implementations, the fibers bring the pulses of radiation 317 and 318 directly to an interior of the chamber 340 through an opening formed in a wall of the chamber 340. Returning to the example of FIG. 3A, the amplified light beam 316 from the drive laser system 315 is reflected from the optical element 322 and propagates through the focusing assembly 326. The focusing assembly 326 focuses the amplified light beam 316 onto the target location 342. The pulses of radiation 317 and 318 pass through the optical element 322 and are directed through the focusing assembly 326 to the chamber 340. Referring to FIGS. 3B-3D, each of the amplified light beam 316, the pulse of radiation 317, and the pulse of radiation 318 are directed to different locations along the “x” direction in the chamber 340. Referring also to FIGS. 3B-3D, a top view of the target material supply apparatus 347 releasing a stream of target material droplets in the “x” direction toward the target location 342 is shown. The stream includes droplets 348a and 348b. The target location 342 is a location that receives the amplified light beam 316 and also can be at the focal point of the EUV collecting optic 346. FIG. 3B shows the chamber 340 at a time t=t1, FIG. 3C shows the chamber 340 at a time t=t2 that occurs after t=t1, and FIG. 3C shows the chamber 340 at a time t=t3, which occurs after t=t2. Each of the amplified light beam 316 and the pulses of radiation 317 and 318 are directed toward different locations along the “x” direction in the chamber 340 at different times. This allows a target material droplet to be converted into a target before reaching the target location 342. FIGS. 3B-3D show an example of a target material droplet (the target material droplet 348a) being converted into the target 55. A time=t1 (FIG. 3B), the pulsed beam of radiation 317 irradiates the target material droplet 348a at the time “t1” at location that is displaced in the “−x” direction from the target location 342. The pulsed beam of radiation 317 transforms the target material droplet 348b into the intermediate target 51. At the time=t2 (FIG. 3C), the intermediate target 51 has moved in the “x” direction closer to the target location 342 and arrives at another location that is displaced in the “−x” direction relative to the target location 342. The pulse beam of radiation 318 irradiates the intermediate target 51 and transforms it into the target 55. The target 55 travels in the “x” direction and arrives at the target location 342 without being substantially ionized. In this manner, the target 55 can be a pre-formed target that is formed at a time before the target 55 enters the target location 342. At the time=t3 (FIG. 3D), the amplified light beam 316 irradiates the target 55 to produce plasma that emits EUV light. In the example shown in FIG. 3A, a single block represents the pre-pulse source 324. The pre-pulse source 324 can be a single light source or a plurality of light sources; for example, two separate sources can be used to generate the pulses 317 and 318. The two separate sources can be different types of sources that produce pulses of radiation having different wavelengths and energies. For example, the pulse 317 can have a wavelength of 10.6 μm and be generated by a CO2 laser, and the pulse 318 can have a wavelength of 1.06 μm and be generated by a rare-earth-doped solid state laser. In some implementations, the pulse of radiation 317 can be generated by the drive laser system 315. For example, the drive laser system can include two CO2 seed laser subsystems and one amplifier. One of the seed laser subsystems can produce an amplified light beam having a wavelength of 10.26 μm, and the other seed laser subsystem can produce an amplified light beam having a wavelength of 10.59 μm. These two wavelengths can come from different lines of the CO2 laser. In other examples, other lines of the CO2 laser can be used to generate the two amplified light beams. Both amplified light beams from the two seed laser subsystems are amplified in the same power amplifier chain and then angularly dispersed to reach different locations within the chamber 340. The amplified light beam with the wavelength of 10.26 μm can be used as the pre-pulse 317, and the amplified light beam with the wavelength of 10.59 μm can be used as the amplified light beam 316. Moreover, the amplified light beam 316, the pulse of radiation 317, and the pulse of radiation 318 are all amplified in the same amplifier. For example, the three power amplifiers 181, 182, and 183 (FIG. 1B) can be used to amplify all of the amplified light beam 316, the pre-pulse 317, and the pre-pulse 318. In this implementation, the amplifier can have three seed lasers, one of which is used to generate each of the amplified light beam 316, the pulse of radiation 317, and the pulse of radiation 318. More or fewer seed lasers can be used. Referring to FIG. 4, a flow chart of an example process 400 for generating EUV light is shown. The process 400 can be performed using the light source 100 or the light source 305. A first pulse of radiation is directed toward a target material droplet to form an altered droplet (410). The first pulse of radiation can be a pulse that has an energy that is sufficient to alter a shape of the target material droplet. The first pulse of radiation can have a duration of at least 1 ns, for example, the first pulse of radiation can have a duration of 1-100 ns and a wavelength of 1 μm or 10 μm. In one example, the first pulse of radiation can be a laser pulse that has energy of 15-60 mJ, a pulse duration of 20-70 ns, and a wavelength of 1-10 μm. In some examples, the first pulse of radiation can have a duration of less than 1 ns. For example, the first pulse of radiation can have a duration of 300 ps or less, 100 ps or less, between 100-300 ps, or between 10-100 ps. The first pulse of radiation can be the first pre-pulse 6 (FIG. 1B) or the pulse of radiation 317 (FIGS. 3A-3D). The altered droplet can be the intermediate target 51 (FIG. 1A) that is formed by irradiating the target material droplet 50 with the first pre-pulse 6. The target material droplet 50 can be a droplet of molten metal, such as tin or any other material that emits EUV when converted to plasma. For example, the altered droplet can be a disk of molten tin formed by striking the target material droplet 50 with the first pre-pulse 6. The force of the impact of the first pre-pulse 6 can deform the droplet into a shape that is closer to a disk that expands, after about 1-3 microseconds (μs), into a disk shaped piece of molten metal. In this example, the disk shaped piece can be considered the intermediate target 51. FIGS. 6C and 8C show an exemplary intermediate target 613 that is disk shaped. The altered droplet or intermediate target can take other geometric forms. For example, in implementations in which the first pulse of radiation is less than 1 ns in duration, the altered droplet can have a shape that is formed from slicing a spheroid along a plane, such as a hemisphere like shape. FIG. 10C shows an exemplary intermediate target 1014 that has a hemisphere shape. In the example shown in FIG. 10C, the intermediate target 1014 is a volume of particles instead of a disk shaped segment of molten tin. A second pulse of radiation is directed toward the altered droplet to form an absorption material (420). The absorption material is the target 55 that receives the amplified light beam and is converted to plasma (by ionization due to the interaction of the amplified light beam with the target 55) that emits EUV light. The second pulse of radiation has energy sufficient to change a property of the altered droplet that is related to absorption of radiation. In other words, striking the altered droplet formed in (420) with the second pulse of radiation changes the ability of the altered droplet to absorb radiation, such as light. Further, the property related to absorption of radiation is changed such that the absorption material is able to absorb a higher portion of incident radiation than the altered droplet. The second pulse of radiation can have a duration of at least 1 ns and an energy of 1-10 mJ. For example, the second pulse of radiation can have a duration of 10 ns and an energy of 5 mJ. The second pulse of radiation can have a wavelength of 1.06 μm. The second pulse of radiation can be the second pre-pulse 7 (FIG. 1B) or the pulse of radiation 318 (FIGS. 3A-3D). Although the energy of the second pulse of radiation can be lower and/or the pulse duration can be longer than a pre-pulse applied directly to the target material droplet, the absorption material (such as the target 55) has physical properties that make the target 55 favorable for generating EUV light. In one example, the intermediate target 51 is a disk of molten tin that, as compared to the target material droplet 50, is thinner along a direction of propagation of an incident pulse of radiation. This intermediate target 51 is more easily broken into fragments of target material than the target material droplet 50, and a smaller amount of energy may be needed to fragment the intermediate target 51. In this example, the second pulse of radiation transforms the intermediate target 51 into a cloud of pieces of target material that, taken together or collectively, have a greater surface area of target material in the path of an oncoming pulse of radiation as compared to the target material droplet 50. The greater surface area provides more target material for interaction with an amplified light beam and can lead to increased ionization of the target material and therefore increased EUV light generation. FIG. 7 shows an example of a second pulse of radiation that transforms the intermediate target 51 into a target 55 that includes fragments of target material. In another example, the intermediate target 51 is again a disk of molten tin that is thinner and wider than the target material droplet. In this example, the second pre-pulse irradiates the intermediate target 51 and generates a cloud of electrons and ions (a pre-plasma) close to the surface of the intermediate target that receives the second pulse of radiation. By creating the cloud of electrons and ions at the surface of the intermediate target 51, the second pulse of radiation alters the electron density and/or the ion density of at least a portion of the intermediate target 51. FIG. 5 shows an example of a second pulse of radiation that changes the electron density and/or ion density of at least part of the modified droplet. An amplified light beam is directed to the absorption material (430). The amplified light beam has energy sufficient to ionize and therefore convert target material in the absorption material (the intermediate target 51) into a plasma that emits EUV light. The amplified light beam can be the amplified light beam 8 (FIG. 1B). The target 55 and the waveform 5 discussed above provide examples. FIGS. 5, 7, and 9 show representations of other exemplary waveforms 500, 700, and 900, respectively, for generating a target. FIGS. 6A-6E, 8A-8E, and 10A-10E show energy of the waveforms 500, 700, and 900 being applied to a target material droplet. Referring to FIG. 5, a plot of an example waveform 500 that can be used to convert a target material droplet to a target that emits EUV light is shown. FIGS. 6A-6D show the waveform 500 transforming a target material droplet to the target that emits EUV light. The target of the example of FIG. 5 and FIGS. 6A-6D is a flat disk of molten metal that has a pre-plasma formed at a surface that faces an oncoming amplified light beam. The surface can face the oncoming amplified light beam if it is pointed towards the amplified light beam, even if the surface is not transverse to the direction of propagation of the amplified light beam. The waveform 500 shows a representation of a first pre-pulse 502, a representation of a second pre-pulse 504, and a representation of an amplified light beam 506. In this example, the first pre-pulse 502 has a pulse duration 503 that is 20-70 ns, and an energy of 15-60 mJ. For example, the first pre-pulse 502 can have a wavelength of 1 μm or 10.6 μm. In one example, the pulse duration 503 is 40 ns, and the energy is 20 mJ. The second pre-pulse 504 can have a pulse duration 505 that is 1-10 ns, an energy of 1-10 mJ, and a wavelength of 1.06 μm. In one example, the duration 505 of the second pre-pulse 503 is 10 ns, and the energy of the second pre-pulse is 1 mJ. The first pre-pulse 502 and the second pre-pulse 504 are separated in time by a delay time 508, with the second pre-pulse 504 occurring after the first pre-pulse 502. The delay time 508 is a time that is long enough to allow a target material droplet that is geometrically altered through an interaction with the first pre-pulse 502 to expand to form the intermediate target 51. The delay time 508 can be 1-3 microseconds (μs). The second pre-pulse 504 and the amplified light beam 506 are separated in time by a delay time 509, with the amplified light beam 506 occurring after the second pre-pulse 504. The delay time 509 is long enough to allow the pre-plasma that the second pre-pulse 504 forms at the surface of the disk shaped target to expand. The delay time 509 can be between 10-100 ns or between 1-200 nanoseconds (ns). FIGS. 6A-6E show side views of a target material supply apparatus that releases target material droplets toward a target location 626 at five different times, t1-t5. The target location 626 is a location in a chamber (such as a chamber 340) that receives the amplified light beam 506 and is at the focus of the collecting optics 346 (FIG. 3A) or the mirror 135 (FIG. 1A). FIG. 6A shows the earliest time, t1, and the time increases from left to right, with FIG. 6E showing the latest time, t5. A target material supply apparatus 620 releases a stream of droplets through a nozzle 624. The stream of droplets includes target material droplets 611 and 610, with the target material droplet 610 being released from the nozzle 624 before the target material droplet 610. FIGS. 6A-6E show the target material droplet 610 being transformed into a target 614 that emits EUV light when struck by the amplified light beam 506. Referring to FIG. 6A, the target material droplet 610 is struck by the first pre-pulse 502. As shown in FIG. 6B, the impact of the first pre-pulse 502 geometrically deforms and spreads the target material droplet 610 into an elongated segment of target material 612. The elongated segment 612 can have a shape that is disk-like and the elongated segment 612 can be molten target material. The elongated segment of target material 612 expands spatially as it travels toward the target location 626. The elongated segment of target material 612 expands for 1-3 μs (the delay time 508). Referring to FIG. 6C, at time=t3, which is 1-3 μs after the first pre-pulse 502 strikes the target material droplet 610, the oblong shaped material 612 has expanded into a disk shaped intermediate target 613 as it follows its trajectory toward the target location 626. Referring also to FIG. 6F, the intermediate target 613 has a width 632 and a thickness 630. The thickness 630 of the intermediate target 613 is less than the width. In the example shown in FIGS. 6C and 6F, the width 632 is in the “x” direction and the thickness 630 is in the “y” direction, and the width 632 is along a direction that is transverse to the direction of propagation of the second pre-pulse 504. However, the intermediate target 613 can have other angular placements. For example, as shown in FIG. 6G, the intermediate target 613 can be angled 45° relative to the direction of propagation of the second pre-pulse 504. Even when the intermediate target 613 is angled relative to the path of the second pre-pulse 504, a thickness 631 of the intermediate target 613 measured along the direction of propagation of the pre-pulse 504 is less than the width of the intermediate target 613. As such, an oncoming light beam (such as the second pre-pulse 504) encounters less target area along a direction of propagation than along a plane that is perpendicular to the path that the oncoming light beam would travel if it went directly through the target material droplet 610. Referring also to FIG. 6D, the interaction between the second pre-pulse 504 and the intermediate target 613 forms a target 614. The interaction creates a pre-plasma 615 that is close to a bulk target material 616. The bulk target material 616 can be target material and can be molten metal. The pre-plasma 615 is allowed to expand over the delay time 509, and the expanded plasma 615 and the bulk target material 616 form the target 614. At time t4, the target 614 arrives at the target location 626. In greater detail, the second pre-pulse 504 impinges on a surface of the intermediate target 613 and heats the surface to form the pre-plasma 615. Because the intermediate target 613 is shaped like disk with the thin dimension presented to the pre-pulse 504, the pre-plasma 615 can utilize a higher portion of the target material in the bulk material 616. After the pre-plasma has expanded for 1-200 ns the pre-plasma and the bulk target material 616 are collectively called the target 614. The amplified light beam 8 arrives at the target 614 before the pre-plasma 615 blows off or dissipates. For example, the amplified light beam 8 can arrive 10-100 ns or 1-200 ns after the second pre-pulse 504 strikes the intermediate target 613. Because the pre-plasma 615 is present when the amplified light beam 506 arrives, the amplified light beam 506 encounters the pre-plasma 615 prior to reaching the underlying bulk target material 616. Compared to the underlying bulk target material 616, the pre-plasma 615 is less reflective and absorbs the amplified light beam 506 more readily. Thus, the presence of the pre-plasma 615 allows a larger portion of the amplifying light beam 506 to be absorbed. Further, in the absence of the pre-plasma 615, the amplified light beam 506 impinges on the bulk target material 616 directly. In this instance, the amplified light beam 8 would encounter a metal surface and would mostly be reflected, with a small amount of the amplified light beam 8 being absorbed to ablate the surface of the bulk target material 616 and form a pre-plasma cloud near the surface. The cloud can be formed 5-20 ns after a pulse impinges on the surface. However, many pulses that have energy sufficient to convert the target material to plasma that emits EUV light have a steep leading edge in the first 10-20 ns of the pulse. The amplified light beam 506 has a leading edge 510 (FIG. 5). The intensity of the leading edge 510 (the portion of the pulse that reaches the target surface over the first 10-20 ns of interaction between the pulse and the target) increases rapidly as a function of time, and increases before the cloud of electrons and ions has had a chance to form and before the heating and evaporation process begins. Thus, without the pre-plasma 615, much of the energetic leading edge 510 of the amplified light beam 8 would be reflected and largely unused. However, the pre-plasma 615 absorbs a portion of the energy in the leading edge 510 and converts it to heat that ablates the bulk target material 616. Referring to FIG. 6E, the amplified light beam 506 converts most or nearly all of the pre-plasma 615 and the bulk target material 616 into EUV light 618. Referring to FIG. 7, a plot of another exemplary waveform 700 that can be used to convert a target material droplet to a target that emits EUV light is shown. FIGS. 8A-8E show the waveform 700 transforming a target material droplet to the target that emits EUV light. The target of the example of FIG. 7 and FIGS. 8A-8E is a collection of fragmented target material. The waveform 700 shows a representation of a first pre-pulse 702, a representation of a second pre-pulse 704, and representation of a an amplified light beam 706. The first pre-pulse 702 has a pulse duration 703 that is 20-70 ns, and an energy of 17-60 mJ. The first pre-pulse 702 can have a wavelength of 1 μm or 10.6 μm. In one example, the pulse duration 703 is 40 ns, and the energy is 20 mJ. The second pre-pulse 704 has a pulse duration 705 that is 1-10 ns and an energy of 1-10 mJ. The second pre-pulse 704 has a wavelength of 1.06 μm. In one example, the duration 705 of the second pre-pulse 703 is 10 ns, and the energy of the second pre-pulse is 5 mJ. In another example, the duration 705 of the second pre-pulse 703 is 10 ns, and the energy of the second pre-pulse is 10 mJ. The first pre-pulse 702 and the second pre-pulse 704 are separated in time by a delay time 708, with the second pre-pulse 704 occurring after the first pre-pulse 702. The delay time 708 is a time that is long enough to allow a target material droplet that is geometrically deformed by the first pre-pulse 702 to expand to form an disk shaped intermediate target. The delay time 708 can be 1-3 microseconds (μs). The second pre-pulse 704 and the amplified light beam 706 are separated in time by a delay time 709, with the amplified light beam 706 occurring after the second pre-pulse 704. The delay time 709 is long enough to allow the fragments that the second pre-pulse 704 forms to disperse to an optimal distance. The delay time 709 can be 100 nanoseconds (ns) to 1 microsecond (μs). Referring to FIGS. 8A-8E, five snap shots of the target material supply apparatus 620 are shown, with time increasing from FIG. 8A on the left to FIG. 8E on the right. FIGS. 8A-8C produce the disk-shaped intermediate target 613 as discussed with respect to FIGS. 6A-6C. FIG. 8D shows the generation of the target 814. The target 814 is a collection of pieces or particles of target material that is formed by irradiating the intermediate target 613 with the second pre-pulse 704. The impact of the second pre-pulse 704 breaks the intermediate target 613 into many fragments of target material, each of which is smaller than the intermediate target 613. Breaking the intermediate target 613 into the fragments provides more target material for the amplified light beam 706 because, collectively, the fragments present more surface area of target material for conversion to plasma. Moreover, because of the thinness of the intermediate target 613, the second pre-pulse 704 may be relatively less energetic and/or longer in duration than a pre-pulse capable of transforming the target material droplet 610 into a collection of fragments. The target 814 arrives in the target location 626 and receives the amplified light beam. EUV light 818 is produced. Referring to FIG. 9, a plot of another exemplary waveform 900 that can be used to convert a target material droplet to a target that emits EUV light is shown. FIGS. 10A-10E show the waveform 900 transforming a target material droplet to the target that emits EUV light. The target of the example of FIG. 9 and FIGS. 10A-10E is a pre-plasma that is formed close to a hemisphere shaped target. The waveform 900 shows a representation of a first pre-pulse 902, a representation of a second pre-pulse 904, and a representation of an amplified light beam 906. The first pre-pulse 902 has a pulse duration 903 that is less than 1 ns. For example, the first pre-pulse 902 can have a wavelength of 1.06 μm, a pulse duration of 300 ps or less, and an energy of 1 mJ-10 mJ. In another example, the first pre-pulse has a duration of 100 ps-300 ps, a wavelength of 1.06 μm, and an energy of 1 mJ-10 mJ. In yet another example, the first pre-pulse 902 has a duration of 150 ps, a wavelength of 1.06 μm, and an energy of 5 mJ. The second pre-pulse 904 has a pulse duration 905 that is 1-10 ns and an energy of 1-10 mJ. The second pre-pulse 904 has a wavelength of 1.06 μm. In one example, the duration 905 of the second pre-pulse 903 is 10 ns, and the energy of the second pre-pulse is 5 mJ. In another example, the duration 905 of the second pre-pulse 903 is 10 ns, and the energy of the second pre-pulse is 10 mJ. The first pre-pulse 902 and the second pre-pulse 904 are separated in time by a delay time 908, with the second pre-pulse 904 occurring after the first pre-pulse 902. The delay time 908 is a time that is long enough to allow a target material droplet that is geometrically deformed by the first pre-pulse 902 to expand to form a hemisphere shaped target. For example, the delay time 908 can be about 1000 ns. The delay time 909 is long enough to allow the pre-plasma that the second pre-pulse 904 forms at the surface of the hemisphere shaped target to expand. The delay time 909 can be 10-100 nanoseconds (ns) or 1-200 ns. Referring to FIGS. 10A-10E, five snap shots of the target material supply apparatus 620 are shown, with time increasing from FIG. 10A on the left to FIG. 10E on the right. The first pre-pulse 902 irradiates the target material droplet 610 to form a hemisphere shaped volume 1012. The hemisphere shaped volume 1012 is a mist or collection of particles 1013 that are distributed throughout a hemisphere shaped space. The particles 1013 are distributed with a density distribution that is minimum at a surface 1002 that faces toward the second pre-pulse 904. The direction of increase of the density distribution contributes to an increased amount of light being absorbed by the volume 1012 because most of the light is absorbed by the volume 1012 before the light reaches a plane of high density that could reflect the light. The hemisphere shaped volume 1012 expands over the delay time 909 to form the hemisphere shaped intermediate target 1014. The second pre-pulse 904 irradiates the hemisphere shaped intermediate target 1014 to generate a pre-plasma at an edge of the intermediate target 1014 and also converts at least some of the particles 1013 into the pre-plasma. Because the particles 1013 are small, it is relatively easy to generate a pre-plasma from the particles 1013. The pre-plasma expands over the delay time 909 to form the target 1015. The target 1015 includes a hemisphere shaped volume 1017 and a pre-plasma 1016. The amplified light beam 906 irradiates the target 1015 to generate EUV light. The pre-plasma 1016 provides a medium that absorbs the amplified light beam 906 readily, thus, the pre-plasma 1016 can enhance and improve the conversion of the amplified light beam into EUV light. Other implementations are within the scope of the following claims. For example, the disk shaped intermediate target 613 can have a shape that is similar to a disk or that includes an indentation in one of the surfaces. Any of the waveforms 5, 500, 700, and 900 discussed above can have more than two pre-pulses that interact with target material. |
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059129336 | summary | TECHNICAL FIELD The present invention relates generally to methods for evaluating nuclear power core operation for Boiling Water Reactors, and more particularly to an improved method and apparatus for determining an operating limit minimum critical power ratio (OLMCPR) so as to effectuate increased efficiency and operation of Boiling Water Reactors. BACKGROUND In a Boiling Water nuclear Reactor there are basically three modes of heat transfer that must be considered in defining thermal limits for the reactor: (i) Nucleate boiling, (ii) transition boiling and (iii) film boiling. Nucleate boiling is the preferred efficient mode of heat transfer in which the BWR is designed to operate. Transition boiling is manifested by an unstable fuel rod cladding surface temperature which rises suddenly as steam blanketing of the heat transfer surface occurs, then drops to the nucleate boiling temperature as the steam blanket is swept away by the coolant flow, then rises again. At still higher fuel rod/bundle operating powers, film boiling occurs which results in higher fuel rod cladding temperatures. The cladding temperature in film boiling, and possibly the temperature peaks in transition boiling, may reach values which could cause weakening of the rod cladding and accelerated corrosion. Thus, fuel rod overheating is conservatively defined as the onset of the transition from nucleate boiling to film boiling. Accordingly, the conventional basis for reactor core and fuel rod design is defined such that some "margin", accommodating various design and operational "uncertainties", is maintained between the most limiting operating condition and the transition boiling condition at all times for the life of the core. The onset of transition boiling can be predicted by a correlation to the steam quality at which boiling transition occurs--called the "critical quality". Steam quality can be readily measured and is generally a function of measuring distance above the boiling boundary (boiling length) for any given mass flow rate, power level, pressure and bundle flow geometry among other factors. A "critical power" is defined as that bundle power which would produce the critical quality of steam. Accordingly, a "critical power ratio" (CPR) is then defined as the ratio of the critical power to the bundle operating power at the reactor condition of interest and is descriptive the relationship between normal operating conditions and conditions which produce a boiling transition. Consequently, the CPR is conventionally used as the figure of merit for rating reactor design and operation. To assure a safe and efficient operation of the reactor, the CPR must be kept above a prescribed value for all of fuel assemblies in the core. Consequently, reactor operating limits are conventionally defined in terms of the most limiting fuel assembly in the core--defined as the "minimum critical power ratio" (MCPR). Reactor operating limits are thus often stated in terms of MCPR. In nuclear power generation engineering, it is widely recognized that there is a possibility, however small, that the occurrence of a reactor transient event combined with the various "uncertainties" and tolerances inherent in reactor design and operation may cause transition boiling to exist locally for some period of time. Accordingly, MCPR operating limits are conventionally set in accordance with the United States Nuclear Regulatory Commission (USNRC) design basis requirement that transients caused by single operator error or single equipment malfunction shall be limited such that, taking into consideration uncertainties in the core operating state, more than 99.9% of the fuel rods are expected to avoid boiling transition. Accordingly, a safety limit minimum critical power ratio (SLMCPR) is defined under current USNRC requirements as the MCPR where no more than 0. 1% of the fuel rods are subject to boiling transition. Notwithstanding the above design basis requirements, developments in fuel design, core loading, and reactor operation over the past years have gradually increased the operating limit minimum critical power ratio (OLMCPR) and reduced the operational "margin" conventionally associated with Boiling Water Reactors (BWRs). Several factors have contributed to the reduction in reactor operational margin. For example, the development of 9-by-9 and 10-by-10 fuel rod bundles having smaller rod diameters has reduced the thermal time constant associated with the fuel rods and made the fuel rods more sensitive to power transients. In addition, the conventional use of a "one-dimensional" power shape model in the mathematical modeling and analysis of the transient response of fuel rods has the effect of further reducing the calculated operating margin. As a result, the OLMCPR has increased to within a range of 1.3 to 1.4--which for most BWRs is typically set from an observance of "fast" pressurization transients (e.g., such as those resulting from a turbine "trip" without bypass). However, contemporary movements toward the use of high energy cores--characterized by power up-rates, long cycles and high capacity factors--necessitates increased critical power ratio (CPR) margins in order to optimize the fuel cycle economy. Moreover, as a result of contemporary optimizations in fuel bundle design, recent increases in operational safety limits translate into a corresponding increase in the operating limit CPR. Motivated by these and other concerns, the inventors of the present invention were led to examine more closely some of the processes conventionally used in evaluating BWR designs and calculating OLMCPR. As a consequence, it was realized that the conventional processes were laden with excessive conservatism that resulted in inaccurate evaluations of reactor performance and calculation of the OLMCPR. For example, the following is a brief list summarizing five of the somewhat more prominent factors identified by the inventors as contributing to excessive conservatism in conventional BWR performance evaluations: i. The use of one-dimensional (1-D) instead of three-dimensional (3-D) methods. PA1 ii. The inconsistent use of radial power shapes. PA1 iii. The addition of "uncertainties" instead of a statistical combination. PA1 iv. A failure to consider direct moderator heating in the correlation for boiling transition. PA1 v. Overly conservative safety limit parameters. The conventional 1-Dimensional modeling methods used for evaluation of transients fail to incorporate the "flattening" of the shape of the radial power distribution that generally occurs during a transient --thus leading to an over prediction of the transient change in critical power ratio (DCPR). PA2 The safety limit is calculated in 3-D using the flattest possible steady state radial power shape in order to maximize the number of rods close to boiling transition. Conventionally, the transient CPR (DCPR) is calculated in 1-D assuming a highly peaked radial power shape in order to drive the bundle to the safety limit. If the same peaked radial power shape that was used for the DCPR evaluation is used for the safety limit, a lower safety limit would result. Conversely, if the same flattened radial power shape used for the safety limit were applied to the transient the DCPR would be reduced. For either situation the OLMCPR would be reduced and operating margin would be increased. PA2 Values representing uncertainties in the calculations for both safety limit and DCPR are currently added linearly. Since the parameters contributing to these "uncertainties" are statistically independent, a "propagation of error" or equivalent method is a more appropriate approach for combining these values. PA2 In an actual fuel bundle, a small fraction of the energy is deposited directly into the fluid. However, in conventional fuel rod modeling for a test bundle, all power is considered as residing entirely in the fuel rods. PA2 Smaller values for "uncertainties" than those conventionally used can be justified--resulting in a lower safety limit. Conventionally, a SLMCPR has been statistically evaluated by using steady state calculations. (See for example, the General Electric publication General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977). A statistical evaluation process similar to the one described in this publication was used in the safety evaluation of the licensing topical report for conventional one-dimensional simulation methods, but was only applied to the evaluation of the transient DCPR. (See the General Electric Publication Qualification of the One-dimensional Core Transient Model for Boiling Water Reactors (Volume 1), NEDE-24154-A, Class I, August 1988). Based on the above considerations, the inventors of the present invention realized that due to the excessive degree of conservatism inherent in the conventional basis used for evaluation of BWR operations, a substantial increase in the operating margin for a BWR could be realized by using a less conservative approach toward determining the OLMCPR--as long as such an approach could be demonstrated as mathematically sound. A resultant benefit is that any substantial increase in the permissible operating margin for a reactor translates into increased operating efficiency, greater fuel generation and/or lowered fuel exhaustion. For example, demonstrating that a particular reactor or reactor design actually has a greater operational margin than may have been previously realized--for example, due to the use of an unnecessarily overly conservative evaluation method--could permit operation at increased output power levels or at comparable power output levels using less fuel. Accordingly, a less conservative and more mathematically sound evaluation method that results in a substantially greater operational margin for Boiling Water Reactors is presented herein for calculating and demonstrating the OLMCPR. Moreover, in principle, the improved method and system of the present invention may also be applied toward the statistical determination of operating limits for other reactor parameters important to reactor safety, for example, the limiting linear heat generation rate (LHGR) in a BWR, the critical heat flux limit in a PWR or LMCR, or the maximum fuel temperature, or the maximum fuel cladding temperature limit in any reactor containing fuel enclosed by a metallic material. These other applications would require that the probability distribution be characterized for each parameter of interest either in the form of a histogram or by some other means such as the generation of a response surface. Briefly, the improved method of the present invention is based on producing a histogram of the number of reactor fuel rods susceptible to operation at "boiling transition" temperatures over a range of variations in selected parametric quantities that are indicative of reactor design constraints and operating conditions. In addition, a core operational modeling approach using multi-dimensional analysis is employed for simulating BWR thermal hydraulics and neutron kinetics during an "anticipated operational occurrence"or AOO in the reactor (for example, a operational occurrence that causes a brief power transient). Basically, in the present invention, all model and reactor plant parameters that may affect the number of rods subject to boiling transition (NRSBT) are first evaluated simultaneously using multi-dimensional modeling of a plurality of fuel rods during a reactor transient condition. The NRSBT is then evaluated statistically in order to determine the OLMCPR directly without the need for first calculating a value for the SLMCPR. Using this approach, the present invention achieves a direct evaluation of OLMCPR of the reactor from a statistical analysis of histograms for the transient condition--as opposed the conventional "indirect" approach of determining an OLMCPR from a combination of values obtained by separately evaluating both a steady state Safety Limit Minimum Critical Power Ratio (SLMCPR) and the change in the critical power ratio (DCPR) due to a transient operational occurrence. Although the use of statistical processes for the evaluation of the "uncertainties" in the modeling of peak cladding temperature for a loss-of-coolant accident has been previously proposed and published by the United States Nuclear Regulatory Commission (USNRC) (See for example, Quantifying Reactor Safety Margin, Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large Break Loss-of-Coolant Accident, NUREG/CR-5249, October 1989 and U.S. Nuclear Regulatory Commission, Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, May 1989), these publications do not define how component uncertainties can be ultimately combined nor how the resulting total uncertainties are to be applied. Moreover, none of the above publications describe or suggest a direct statistical evaluation of the NRSBT during the transient. DISCLOSURE OF THE INVENTION In one aspect, the present invention is an improved method for determining the operating limit minimum critical power ratio (OLMCPR) of a Boiling Water Reactor (BWR) nuclear core. This improvement results in the realization of greater operating margins for the core which in turn results in more efficient and cost effective core operation and/or configurations. In this regard, the present invention also provides an improved and much more direct approach toward illustrating compliance of a nuclear reactor with USNRC licensing requirements than the overly conservative processes conventionally used for such purposes. In accordance with another aspect, the present invention is a data processing system including a computer having memory and various I/O or display devices that is specifically programmed for providing simulation of transient operational events in a BWR and for a subsequent compilation and display of one or more response histogram(s) that incorporate all the inherent "uncertainties" associated with reactor plant initial state conditions and other parameter(s) of interest or importance. Basically, a reactor power transient condition is first simulated and evaluated for numerous fuel rods (or rod bundles) simultaneously using a multi-dimensional modeling technique on a computer. After simulating and evaluating the reactor power transient condition--such as might be caused by single operator error or single equipment malfunction--the expected Number of fuel Rods Subject to Boiling Transition (NRSBT) is determined by calculating and summing probability densities for the CPR values associated with individual fuel rods or fuel rod groups. The probabilities determined for each rod depend on the uncertainty in the critical power correlation and other "uncertainties" which include but are not necessarily limited to uncertainties in reactor plant parameters, initial conditions, power distribution, process data, fuel bundle design, manufacturing, modeling methods, and model inputs. Consequently, such calculations are performed over a sufficient number of different iterations wherein all major modeling and plant state parameters are "perturbed"--for example, using a conventional Monte-Carlo statistical analysis approach--such that an NRSBT distribution histogram is established which statistically incorporates all such uncertainties. This may be performed for any number of selected transient criteria and the resultant histogram(s) is (are) further statistically analyzed and displayed by the data processing system along with relevant selected statistical data. In this manner, compliance or non-compliance of a particular BWR design can be statistically determined and demonstrated for any number of selected transient criteria. The NRSBT distribution histogram is analyzed using statistical methods to determine the "central tendency" of the distribution. Typically the mean or median is used as a statistic to quantify central tendency. The value of this statistic is defined here as the nominal value. In the discussions that follow, examples will be given where the mean value is chosen as the nominal value although the present invention is not limited to this choice. Use of the median value or the value of some other statistic for central tendency as the nominal value is also contemplated as part of the present invention. The uncertainty in the nominal value of the statistic that is used to quantify central tendency is expressed in terms of a "confidence interval" for the nominal value. A confidence interval is defined such that there is a specified probability (usually of 50% or greater) that the interval contains the nominal value. For example, a 95% probability that the interval bounds the mean, defines a 95% confidence interval for the mean. The specified probability used to establish this confidence interval is called the "level of confidence" or confidence level. For the present invention the susceptibility to boiling transition during the transient is quantified statistically as either (1) the probability that a single rod in the core is susceptible to boiling transition or (2) the expected fraction of total rods in the core susceptible to boiling transition. Such a statistical relationship is possible because each individual trial value of NRSBT has been determined by summing the probabilities that individual fuel rods have CPR values less than 1.0 during the transient. The nominal value for each NRSBT distribution can also by the present invention be associated with the distribution of initial rod CPR values for all fuel rods in the core. It is by this process that the present invention can establish the relationship between the minimal initial MCPR value for all fuel rods in the core and the probability and confidence level that the fuel rods will be susceptible to boiling transition during the transient. The minimal initial MCPR value for the core when determined in this way using the probability and confidence level established by the USNRC design basis requirement for the number of rods not susceptible to boiling transition during the AOO transient, is by definition the minimum Operating Limit MCPR required to demonstrate compliance. Although the improved methods, as described herein below, are preferably implemented using a high speed data processing system capable of processing simulation routines that require highly accurate calculations and multiple reiterations, the present invention is not intended as limited to any one particular type of computer or data processing system. Any generic data processing system having sufficient, speed, storage memory and programmable computational capabilities for implementing statistical data analysis/reduction may be utilized to implement the present invention. |
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description | 1. Field of the Invention The present invention relates to a semiconductor processing apparatus and a semiconductor processing method which irradiate energy beams onto a semiconductor film provided on a substrate to crystallize the semiconductor film by a heat treatment. 2. Description of the Related Art Flat type display devices, such as a liquid crystal display device and an organic EL (Electro Luminescence) display device, use thin film transistors (Thin Film Transistors: TFTs) as switching elements for effecting active matrix display of a plurality of pixels. Thin film transistors include a polycrystalline silicon TFT which uses a polysilicon (poly-Si), microcrystalline silicon (μc-Si) or the like in an active region, and an amorphous silicon TFT which uses an amorphous silicon (amorphous Si) in an active region. Of the two types of TFTs, the polysilicon TFT has a feature that the carrier mobility is greater by ten to 100 or so times as compared with the amorphous silicon TFT, and thus has an excellent characteristic as a material for a switching element. Because existing pixel switching elements for driving a liquid crystal do not need such a high carrier mobility, amorphous silicon TFTs are generally used as pixel switches. It is however desirable to use switching elements which have a higher carrier mobility than that of the amorphous silicon TFT in organic EL or high frame usage which is expected to be used in the next generation TFTs. However, the characteristics of the polycrysilicon TFT, such as carrier mobility and threshold voltage, are likely to vary from one device to another due to the presence of both a crystal region and a grain boundary region in a channel region which is the core of the transistor. It is said that the carrier mobility and the device variation have a trade-off relation such that as the carrier mobility becomes greater, the device variation becomes larger. Multiple studies have been done on such a relation originating from the unique grain boundary of the polysilicon TFT. For example, JP-A-2004-87535 (Patent Document 1) discloses a technique of irradiating an excimer laser beam over a diffraction grating mask in addition to a control method for the crystal orientation in repetitive irradiation. That is, positional control on the grain boundary is carried out with the diffraction grating mask in such a way that crystal growth occurs at a point of the intensity-modulated minimum light intensity as the origin, and the position of the grain boundary is set to a point of the maximum light intensity. As another technique, a scheme of crystallization with laser scanning is disclosed. For example, JP-A-2004-87667 (Patent Document 2) discloses SELAX (Selectively Enlarging Laser Crystallization). This method moves (scans) a continuous wave (CW) laser beam or a pseudo CW laser beam with a very high pulse frequency of several tens of MHz or so in one direction in relative to a silicon film deposited on a substrate. The laser scanning causes a crystal to grow in one direction. The grain boundary is formed substantially in parallel to the direction of growth. Therefore, the electric conduction in the direction of crystal growth reduces the density of the grain boundary where carriers cross, thus making it possible to increase the carrier mobility. The closer to a monocrystal the channel portion becomes, the more prominent a difference in in-plane crystal orientation, a slight angular difference in plane orientation, and a difference in internal defect become. It is therefore desirable to form a crystal film of a higher quality of the level of a monocrystalline Si substrate to suppress a device variation. JP-A-2007-281422 (Patent Document 3) discloses a technique of forming a polycrystal with a controlled number of grain boundaries where carriers cross, not making silicon in the channel into a monocrystal. Further, JP-A-2007-281420 (Patent Document 4) discloses a technique of scanning a channel region with a spot laser beam at equal pitches to form grain boundaries. This makes it possible to form a transistor with a smaller device variation while having a relatively high carrier mobility. Those schemes however have a problem of making it difficult to increase the throughput. That is, it is necessary to carry out beam scanning a number of times to crystallize a silicon film in the channel with a beam of a small spot size, and this requirement stands in the way of improving the productivity. Accordingly, it is desirable to provide a technique capable of improving the throughput in forming a semiconductor device through a heat treatment on a semiconductor film using an energy beam. According to an embodiment of the present invention, there is provided a semiconductor processing apparatus including a stage on which a substrate having a semiconductor film to be processed is to be mounted, a supply section that supplies a plurality of energy beams onto the semiconductor film mounted on the stage in such a way that irradiation points of the energy beams are aligned at given intervals, and a control section that moves the plurality of energy beams and the substrate relative to each other in a direction not in parallel to alignment of the irradiation points of the plurality of energy beams supplied by the supply section, and scans the semiconductor film with the irradiation points of the plurality of energy beams in parallel to thereby control a heat treatment on the semiconductor film. According to another embodiment of the present invention, there is provided a semiconductor processing method including the steps of supplying a plurality of energy beams onto a semiconductor film provided on a substrate to be processed in such a way that irradiation points of the energy beams are aligned at given intervals, and moving the plurality of energy beams supplied and the substrate relative to each other in a direction not in parallel to alignment of the irradiation points of the plurality of energy beams supplied by the supply section, and scans the semiconductor film with the irradiation points of the plurality of energy beams in parallel to thereby control a heat treatment on the semiconductor film. According to the embodiments of the invention, a substrate is scanned with a plurality of energy beams in parallel with the irradiation points of the energy beams aligned at given intervals, thus making it possible to increase the number of regions to be processed at a time as compared with a case of scanning a substrate with a single beam irradiation point. The present invention can therefore improve the throughput in forming a semiconductor device through a heat treatment on a semiconductor film using an energy beam. A preferred embodiment of the present invention is described below with reference to the accompanying drawings. A semiconductor processing apparatus according to the embodiment, which is described herein, is the apparatus that is used in crystallization and activation of an amorphous semiconductor film of silicon, germanium or the like formed on a substrate in the process of crystallization/activation of a so-called semiconductor film. <Schematic Configuration of Semiconductor Processing Apparatus> FIG. 1 is a diagram illustrating the schematic configuration of the semiconductor processing apparatus according to the embodiment. The semiconductor processing apparatus according to the embodiment includes a stage 20 on which a substrate 10 having a semiconductor film 11 to be processed is to be mounted, a supply section 30 that supplies energy beams, and a control section 40 that controls the stage 20 and the irradiation of the energy beams. The stage 20 has a capability of moving the substrate 10 mounted thereon in vertical and horizontal (XY) directions. The stage 20 has a capability of moving the substrate 10 in the altitude (Z) direction and in the rotational (θ) direction as needed. The supply section 30 supplies a plurality of energy beams B onto the semiconductor film 11 mounted on the stage 20 in such a way that irradiation points of the energy beams B are aligned at given intervals. A laser beam is used as the energy beam B. Therefore, the supply section 30 has a light source part which emits a laser beam, a beam shaper which shapes the laser beam emitted from the light source part, and a dividing part which divides the laser beam into a plurality of energy beams B. The control section 40 controls the movement of the stage 20 in the in the XY directions, and in the Z direction and the θ direction as needed, and gives an instruction to the supply section 30 to control the amount of energy of the energy beams B. Accordingly, the state of a heat treatment on the semiconductor film 11 on the substrate 10 is controlled. <Specific Example of Semiconductor Processing Apparatus> FIG. 2 is a configurational diagram illustrating a specific example of the semiconductor processing apparatus according to the embodiment. The semiconductor processing apparatus shown in FIG. 2 includes the stage 20 on which the substrate 10 is mounted, a light source part 31 which generates a laser beam, a laser beam combining part, a beam shaper 32 which shapes the irradiation shape of the laser beam, and a diffraction grating mask irradiator 33 which is a dividing part to subject the laser beam to intensity modulation to divide the laser beam. The light source part 31, the beam shaper 32 and the diffraction grating mask irradiator 33 are provided in the supply section 30 shown in FIG. 1, and those components and the stage 20 are controlled by the control section 40. Of the components, the stage 20 is configured to be provided with a slider capable of keeping the substrate 10 mounted on the stage 20 horizontally, and freely moving/scanning the substrate 10 in the X-Y direction in the horizontal plane. In case where the substrate 10 mounted on the stage 20 has a flexibility and is bent flexibly, the stage 20 may be configured to have a roll-to-roll type roll-up mechanism at an end portion to roll up the substrate 10 in one direction (e.g., plus or minus Y direction) in which the substrate 10 is scanned. The light source part 31 includes a semiconductor laser oscillator. As the semiconductor laser oscillator, one which generates a laser beam having the adequate characteristic and wavelength according to what process is to be executed using the semiconductor processing apparatus is selected. In case where the semiconductor processing apparatus is used in crystallization and activation of a semiconductor film formed of amorphous silicon, for example, a compound semiconductor laser oscillator containing Ga and N and having an oscillation wavelength of 350 nm to 470 nm is used in consideration of the laser stability and the absorption coefficient of the semiconductor film. Further, a semiconductor laser or a solid state laser with the second harmonics of a titanium-saphaire laser, the second harmonics of an Nd:YVO4 laser, the second harmonics of a YAG laser, or the like, or a Kr gas laser or the like as a gas laser is used preferably. The beam shaper 32 can cut out individual regions of a laser beam having a profile approximate to the Gaussian shape, generated from the light source part 31, with a cylindrical or honeycomb lens, and overlay the cut-out regions to shape the laser beam into a top-hat type predetermined beam shape. Specifically, the beam shaper 32 has a mirror which reflects the laser beam emitted from the light source part 31, a lens which transforms the laser beam reflected at the mirror into parallel light, a homogenizer and lens which form the parallel light into a top-hat type laser beam, and a mirror which guides the top-hat type laser beam to the diffraction grating of the diffraction grating mask irradiator 33. Those optical components (which are an optical head) are configured to be compact, and may be formed on a linear drive type slider to be able to move fast at an equal velocity in one axial direction of the substrate 10 as needed. In this case, it is desirable to reduce the load and size of the optical head part by designing the optical head part with optical fibers or the like introduced before the beam combining part except for the case other than using a small-sized semiconductor laser or the like. The laser beam which is shaped into a top-hat type beam shape by the beam shaper 32 is modulated by the diffraction grating of the diffraction grating mask irradiator 33, yielding a plurality of energy beams whose irradiation points are aligned at given intervals. FIG. 3 is an exemplary cross-sectional view illustrating the diffraction grating. The diffraction grating is linear concavoconvexes aligned on one surface of a glass substrate at a predetermined pitch (grating pitch), so that an input semiconductor laser beam is subjected to intensity modulation with an interference pattern. In the embodiment, the grating pitch a is 0.4 to 2.0 μm, for example, and the interference pitch d is a/2. A top-hat type laser beam is modulated to provide alignment of multiple beam irradiation points (beam spots) with the interference pitch. According to the semiconductor processing apparatus of the embodiment, the multiple beam spots are irradiated onto the substrate in parallel. When the optical head side is fixed in use, the stage 20 is configured to move in a direction not in parallel to the pitch direction of the diffraction grating. This makes it possible to provide such setting that the maximum intensity point (beam spot) modulated with the diffraction grating crosses the channel portion of the semiconductor device which is designed on the substrate 10 mounted on the stage 20. In other words, according to the configuration, the substrate 10 is scanned in one direction (e.g., X direction) with the laser beam which forms an image on the substrate 10 while having an intensity distribution. The overall processing speed can be increased by irradiating a plurality of beam spots on the specified channel region on the substrate 10. It is therefore desirable to provide an observation part which can ensure observation of an alignment mark provided on the substrate 10 at a predetermined position. The observation part can be provided on a part other than the optical head as long as the positional relationship with respect to the optical head is maintained. The observation part is a part for observing an alignment mark to be a same reference point marked on the top surface of the substrate beforehand. The observation part includes an illumination light source which irradiates illumination light, a shaping part which shapes the illumination light emitted from the illumination light source, an image sensing part which receives reflected light of the illumination light shaped by the shaping part, and a monitor for showing an image sensed by the image sensing part. A processing unit which includes the stage 20 and the supply section 30 may be increased in quantity within the range where adjustment of the individual processing units is possible. Parallel processing with the individual processing units can improve the throughput further. <Semiconductor Processing Method> A semiconductor device, such as a thin film transistor, is fabricated with the semiconductor processing apparatus having the foregoing configuration in the following procedures. First, the light source part 31 of the supply section 30 generates a laser beam or an energy beam, and laser beams having a plurality of irradiation points aligned at given intervals are irradiated on the substrate via the beam shaper 32 and the diffraction grating mask irradiator 33. To scan the substrate with laser beams with a plurality of irradiation points in this state, the movement of the laser beams relative to the substrate is carried out by driving the stage 20 or driving the optical head. As a result, the semiconductor film 11 on the substrate 10 can be scanned with the laser beams with a plurality of irradiation points aligned at given intervals. Thereafter, scanning with the laser beams in the X direction is repeatedly executed while moving the substrate 10 on the stage 20 in the Y direction. This makes it possible to execute the process of irradiating laser beams at arbitrary positions on the substrate 10 in the two-dimensional (XY) space. According to the semiconductor processing apparatus, the substrate 10 mounted on the stage 20 is relatively scanned with the laser beams, making it possible to scan the substrate 10 with the laser beams at a high speed of, for example, over 0.1 m/sec to effect continuous irradiation of the laser beams on the substrate 10. It is also possible to execute annealing to reduce the thermal load on the substrate 10 by controlling the moving speed of the laser beams with respect to the substrate 10 and the amount of energy of the laser beams. As a result, annealing with the thermal load on the substrate 10 reduced can be carried out by adjusting the moving speed of the laser beams fast or reducing the amount of energy. For example, it is possible to use a plastic substrate or the like with a lower heat resistance in the annealing process for crystallizing the semiconductor film on the substrate 10. This makes it possible to realize a mass-production process with a suppressed equipment cost in the roll-to-roll based equipment using a flexible substrate in the fabrication process for a thin film semiconductor device, for example. There is a case where an electrode pattern or the like is formed on the substrate 10 where the laser beams are irradiated or where the substrate is rolling significantly. When there is not a mechanism of normally optimizing the focus of the laser beams, therefore, irradiation needs to be carried out with a laser having a considerable focus depth. In case of using a laser objective lens with a high convergency, it is possible to previously monitor an irradiation area with a device having a focus servo function and control the annealing laser beam in such a way as to be at an optimal focus position when the laser beam comes to the looked-ahead area. <Direction of Alignment of Irradiation Points of Energy Beams> FIG. 4 is an exemplary diagram illustrating the direction of alignment of the irradiation points of energy beams. According to the embodiment, in irradiating energy beams with a plurality of irradiation points aligned at given intervals on a substrate, the direction of alignment of the plurality of irradiation points is tilted with respect to the scan direction. In the example shown in FIG. 4, the area surrounded by a rectangular frame in the diagram is the laser beam irradiation area formed by a plurality of irradiation points, and the arrow in the diagram shows the scan direction. As the direction of alignment of a plurality of irradiation points is tilted with respect to the scan direction, there is a time lag between the timings at which beam spots hit scanning rows at the respective irradiation points, making it possible to adequately form crescent-shaped crystal grains as shown in FIG. 4. Here, the crystal grains are formed as follows. When the irradiation point (beam spot) of the laser beam is irradiated on the semiconductor film, a part of the semiconductor film or the entire semiconductor film is completely melted with the heat energy. As the semiconductor film is scanned with the beam spot, therefore, solidification progresses with the passing of the laser beam in the scan path of completely melting the semiconductor film, forming a line of crystal grains along the scan center of the laser beam. At this time, the laser beam is made to have, for example, a Gaussian-shaped profile having a distribution with a high center portion and a lower peripheral portion, the temperature of the irradiation portion of the laser beam corresponds to the Gaussian-shaped profile of the laser beam profile (Beam Profile), and becomes highest in the scan center of the laser beam and lowest at both ends thereof. As a laser beam is irradiated while scanning therewith is carried out in the scan direction, crystal solidification starts from positions apart from the scan center (both ends of the scan path of the laser beam) in the scan path of completely melting the semiconductor film, producing a given number of crystal seeds at both ends of the scan path. As the scanning with the laser beam progresses further, solidification progresses in the scan direction toward the scan center with the crystal seeds pulled toward the scan center in the scan direction, and the crystal in the scan center is crystallized last. At this time, the scan speed and output of the laser beam may be adjusted within the range of the foregoing irradiation conditions so that rendezvous of solidification occurs in the scan center. This provides hemicrescent-shaped crystal grains which extend toward both end sides of the scan path from the scan center, i.e., crystal grains which have a shape obtained by bisecting a crescent at a line-symmetrical line. Because scanning with a plurality of irradiation points aligned at given intervals is carried out in parallel in the embodiment, tilting the direction of alignment of the irradiation points with respect to the scan direction allows melting and solidification to be carried out continuously with a time lag for each scan path of adjacent irradiation points, thus forming the crescent-shaped crystal grains. FIG. 5 is an exemplary diagram showing the state of a crystal after laser annealing. According to the embodiment, the laser beam is interfered with the diffraction grating for intensity modulation to thereby form a plurality of irradiation points, so that grain boundaries a corresponding to the number, n, of interference points (beam spots) in association with the interference points. In addition, crescent-shaped crystal grains b are formed between adjacent interference points or between the grain boundaries a. In case of forming a thin film transistor, therefore, the intervals of a plurality of irradiation points, the scan speed thereof and the amount of energy thereof are controlled in such a way that grain boundaries a are periodically provided in the region of the semiconductor film which becomes the channel portion of the thin film transistor in the direction of the channel length. Accordingly, a plurality of the grain boundaries a are formed in the channel portion to ensure the uniform characteristic of the thin film transistor over the range where the carrier mobility is maintained. Specifically, it is desirable to set the conditions in such a way that several tens of (e.g., fifty or so) grain boundaries a in the direction of the channel length. <Regarding Beam Spot> The beam spot of a laser beam which is to be emphasized by the diffraction grating is described below. FIG. 6 is an exemplary diagram illustrating the beam spot size of a laser beam formed by the diffraction grating. In the diagram, Δx is the width of the laser beam incidental to the diffraction grating, λ is the wavelength of the laser beam, and d is the pitch of the diffraction grating. In this case, a distance g for the laser beam to be converged by the diffraction grating is given byg={√(d2−λ2)/2λ}·Δx. Given that θ is the angle of the laser beam perpendicularly incident to the diffraction grating when the laser beam is diffracted and output, the pitch P of the appeared interference pattern emphasized by the diffraction grating is given byP=λ/2 sin θ=λ/2(λ/d)=d/2. The beam spot of the laser beam formed by the diffraction grating has a beam profile approximated to a Gaussian-shaped profile. FIG. 7 is a diagram showing a beam profile based on the relative radiation intensity of a beam. In general, for a beam spot having the profile of the relative radiation intensity approximated to a Gaussian-shaped profile, the size of the beam spot at the position where the peak value of the relative radiation intensity becomes 1/e2 (e is the base of the natural logarithm) or 13.5% is defined as a beam size φ. FIG. 8 is a diagram showing the relation between the beam energy and crystal grains when a semiconductor film is subjected to a heat treatment with a beam size φ of 1.2 μm as defined above. As apparent from the diagram, the size of crystal grains formed by the heat treatment on a semiconductor film with irradiation of a laser beam is substantially less than the beam spot size. According to the embodiment, this property is used to set the interval between, for example, the diffraction grating and the substrate, and the amount of energy so that the beam spots of a plurality of laser beams have a Gaussian shape, thereby adequately controlling the size of the crystal grains formed. FIGS. 9A to 9C are exemplary diagrams showing the beam profile of a laser beam formed by the diffraction grating. As shown in FIG. 9A, the pitch of a plurality of laser beams formed by the diffraction grating becomes d/2 where d is the pitch of the diffraction grating. The above-described distance of the diffraction grating diffraction grating and the substrate causes the beam profile of each laser beam to be approximated to a Gaussian-shaped profile. According to the embodiment, as shown in FIG. 9B, beams are formed in such a way that the beam profile of a plurality of laser beams formed by the diffraction grating is approximated to such a profile that the cross section at any angle of rotation about the axis penetrating the peak value of the relative radiation intensity is approximated to a Gaussian shape. That is, according to the embodiment, at the time of irradiating laser beams with a plurality of irradiation points (beam spots) aligned at given intervals on a substrate, the direction of alignment of a plurality of beam spots is tilted with respect to the scan direction. As each beam spot moves on the scan path along the scan direction at this time, the energy distribution on the scan path becomes equivalent to the cross-sectional shape of the beam profile along a direction perpendicular to the scan direction. Therefore, as beams are formed to have a beam profile which is approximated to such a profile that the cross section at any angle of rotation about the axis penetrating the peak value of the beam profile is approximated to a Gaussian shape, it is possible to normally obtain a beam profile approximated to a constant Gaussian-shaped profile on the scan path even if a predetermined angle is formed between the direction of alignment of beam spots and the scan direction. That is, the peak position of the beam profile on the scan path becomes linear along the scan path, so that the distribution of heat energy on the scan path can be made constant. As shown in FIG. 9C, at the time of irradiating laser beams with beam spots aligned at given intervals and carrying out scanning with the laser beams, a tangent line S which is tangential to two adjacent beam spots on the right and left sides respectively is assumed so as not to produce a gap between the beam spots where a laser beam is not irradiated. The angle θ between the direction of alignment of beam spots and the scan direction (tangent line S) is set in such a way that the tangent line S is in the scan direction. FIG. 10 is an exemplary diagram illustrating an angle between alignment of beam spots and the scan direction. On the assumption of a tangent line S which is tangential to two adjacent beam spots on the right and left sides respectively and is in the scan direction, as mentioned above, the direction of alignment of beam spots is tilted by θ with respect to the scan direction. This prevents a gap where a laser beam is not irradiated from being produced in the scan path formed by the beam spots, making it possible to irradiate the entire laser-beam irradiation area with the laser beam to perform a heat treatment thereon. The tilting of the direction of alignment of beam spots produces a time lag between irradiation positions along the scan direction of the beam spots, making it possible to form crescent-shaped crystal grains b and grain boundaries a. <Method of Dividing Energy Beam> FIGS. 11A and 11B and FIGS. 12A and 12B are exemplary diagrams for explaining methods of dividing a laser beam as an energy beam. Although the foregoing description has given an example of using one diffraction grating as a method of dividing a laser beam having irradiation points aligned at given intervals, the division of a laser beam can be achieved by other methods. The example shown in FIG. 11A utilizes multiple beam interference caused by overlapping two diffraction gratings. In the example shown in FIG. 11B, a laser beam is branched into two optical paths by a half mirror and beam interference is caused by light rays of the two optical paths. The example shown in FIG. 12A has a plurality of optical fibers arranged linearly, so that laser beams are input to the individual optical fibers which output the laser beams from their end faces. This makes it possible to set the intervals of the beam spots based on the pitch of the optical fibers. The example shown in FIG. 12B uses a multiemission laser. The multiemission laser is a laser device which has a plurality of emission points formed on a single substrate. Accordingly, laser beams are emitted from multiple points according to the pitch set at the time of fabrication, and are irradiated on the substrate. In any of the examples, the laser beam and the substrate are moved relative to each other in a direction which not in parallel to the alignment of a plurality of irradiation points of the laser beam. The case where a predetermined angle is provided between the alignment of a plurality of irradiation points and the scan direction can be coped with rotating the optical unit having the aforementioned dividing units or the stage on which the substrate is mounted. <Improvement on Throughput> FIG. 13 is an exemplary diagram illustrating the scan direction in irradiation of laser beams according to the embodiment. In case where a single scan path row N is formed by the alignment of a plurality of laser beams, first, scanning is carried out in the order of arrows A1, A2, A3, A4 and A5 in this embodiment. Next, in scanning the next row, the scanning direction is turned back from the arrow A5, and then scanning is carried out in the opposite direction to the previous scan direction in the order of arrows B1, B2, B3, B4 and B5. Further, in scanning the next row, the scanning direction is turned back from the arrow B5, and then scanning is carried out in the opposite direction to the previous scan direction in the order of arrows C1, C2, C3, C4 and C5. As apparent from the above, when scanning of one row is completed, the scanning direction is turned back in the opposite direction for a next row to carry out scanning in order. It is not therefore necessary to return to the same position for each scan row, minimizing the moving distance of the optical head or the stage. This can improve the throughput at the time of scanning. FIG. 14 is an exemplary diagram showing the state of crystal grains in case of effecting turn-back scan. In scanning a single scan path row N, the scan direction in the direction of the arrow A is opposite to the scan direction in the direction of the arrow B, so that the direction of the crescent-shaped crystal grains is reversed. Further, the scan direction in the direction of the arrow B is opposite to the scan direction in the direction of the arrow C, so that the direction of the crescent-shaped crystal grains is reversed. As apparent from the above, the direction of the crescent-shaped crystal grains is alternately reversed according to the scan direction. <Fabrication Method for Thin Film Semiconductor Device> The following describes a fabrication method for a thin film semiconductor device, which is carried out following the above-described crystallization method. A fabrication method for a semiconductor device which has a plurality of thin film transistors TFT provided on the same substrate is described below. The associated diagrams mainly show only a single thin film transistor forming portion. First, as shown in FIG. 15A, the entirety of each active region 3a set on a semiconductor film 3 on a substrate 1 is selectively crystallized by the above-described crystallization method. Then, band-line crystal grains b are formed in each active region 3a in such a way as to cross the active region 3a. Accordingly, grain boundaries a with a width W1 (=several hundred nm) therebetween are periodically aligned in such a way as to cross the active region 3a. At this time, the length of the belt-like crystal grains b is set to several μm to several hundred μm or so in conformity with the standards for thin film transistors. Next, as shown in FIG. 15B, the semiconductor film 3 is patterned and etched in a predetermined pattern to leave the crystallized active regions 3a, and each active region 3a is divided into islands of a predetermined shape for device isolation. In this case, as illustrated, the semiconductor film 3 may be patterned and etched so that the semiconductor film 3 portion which is not crystallized does not remain around the active region 3a. Alternatively, the semiconductor film 3 may be patterned and etched so that the semiconductor film 3 portion which is not crystallized remains around the active region 3a. In this case, the entire crystallized region in the island-patterned region becomes the active region, while an amorphous region remaining therearound becomes an isolation region. Such patterned etching of the semiconductor film 3 may be carried out before the crystallization process. In this case, the crystallization process is performed on each island-patterned semiconductor film 3 including a region reserved for the active region 3a. Next, a gate insulating film (not shown) is formed on an upper portion of the substrate 1, with the patterned active region 3a covered. This gate insulating film may be formed of silicon oxide or silicon nitride, and can be formed by ordinary PE-CVD (Plasma-Enhanced Chemical Vapor Deposition) known, as well as known SOG (Spin on Glass) which forms an applied insulating film. The formation of the gate insulating film may be carried out before the patterned etching of the semiconductor film 3. Next, as shown in FIG. 16, a gate electrode 5 with a shape to cross the center portion of each island-divided active region 3a is formed on the gate insulating film. It is important to form the gate electrode 5 along the extending direction of the grain boundaries a (the extending direction of the belt-like crystal grains b). FIG. 17 shows in enlargement the part A in FIG. 16. As shown in those diagrams, the gate electrode 5 is provided so as to cross that portion of the active region 3a which is designed to have a predetermined width W, the width of the active region 3a where the gate electrode 5 crosses becomes the channel width W. That is, the grain boundaries a are provided in such a way as to cross the channel portion C under the gate electrode 5 in the direction of the channel width W. It is assumed that the line width of the gate electrode 5 (which corresponds to the channel length L) is designed based on the standards for the thin film transistors to be fabricated here, and a predetermined number of grain boundaries a are arranged so as to cross the channel portion C in the direction of the channel width W. For thin film transistors with the same characteristics, it is important to provide approximately the same number of grain boundaries a at the channel portion C. The “approximately the same number” is preferably in a range of plus or minus one to a predetermined number. The less a variation in the actual number of the grain boundaries a to be provided at the channel portion C with respect to a predetermined number, the more uniform a variation in the characteristics of thin film transistors can be made. It is therefore desirable that the number of the grain boundaries a to be provided at the channel portion C be two or more, and it is better to set the number larger. Specifically, as will be described in the later description of the embodiment, it is preferable to set the width, W1, of the crystal grains b (i.e., the pitch of the grain boundaries) according to the channel length L in such a way that 25 or so grain boundaries a extending in the direction of the channel width W are provided at the channel portion C. It is to be noted however that the greater the grain boundaries a at the channel portion C which cross the direction of the channel length L, the lower the carrier mobility in the direction of the channel length L, so that it is better to set the number of the grain boundaries a as large as possible within the range where the carrier mobility is kept to a certain high level. In addition, it is important to form the gate electrode 5 in a predetermined state with respect to the grain boundaries a provided at each active region 3a as mentioned above. In the previous crystallization process, therefore, as shown in FIG. 18, the scan direction of a laser beam in each active region 3a is set according to the layout direction of the gate electrode 5, so that the extending directions of the belt-like crystal grains b and the grain boundaries a match with the layout direction of the gate electrode 5. At the time of forming the gate electrode 5, first, an electrode material layer of, for example, aluminum is deposited by sputtering or vapor deposition, and then a resist pattern is formed on the electrode material layer by lithography. Thereafter, with the resist pattern used as a mask, the electrode material layer is etched to pattern the gate electrode 5. The formation of the gate electrode 5 is not limited to such procedures, and may be carried out by a scheme of applying and printing, for example, fine metal particles. In etching the electrode material layer at the time of forming the gate electrode 5, the gate insulating film may be etched subsequently. Next, as shown in the cross-sectional view of FIG. 19, a source/drain 7 having an impurity implanted into the active region 3a in a self-aligned manner is formed by ion implantation with the gate electrode 5 used as a mask and the subsequent annealing process. The cross section in FIG. 19 corresponds to the cross section along direction X-X′ in FIG. 16. As a result, the channel portion C is formed by an impurity-free portion in the crystallized active region 3a under the gate electrode 5. The source/drain 7 and the channel portion C under the gate electrode 5 are formed of polycrystalline silicon yielded by crystallizing the semiconductor film 3, thus yielding a thin film semiconductor device 100 in which a plurality of top-gate type thin film transistors TFT (i.e., polysilicon TFT) using a polysilicon thin film are provided on the same substrate 1. In case of preparing a liquid crystal display device, for example, as a display device using such thin film transistors TFT as switching elements, the following processes are performed. First, as shown in FIG. 20A, an interlayer insulating film 210 is formed on the substrate 1 of the thin film semiconductor device 100, covering the thin film transistor TFT. Next, contact holes 210a which reach the source/drain 7 of the thin film transistor TFT are formed in the interlayer insulating film 210. Then, wirings 230 connected to the source/drain 7 via the contact holes 210a are formed on the interlayer insulating film 210. Next, a planarized insulating film 250 is formed, covering the wirings 230, contact holes 250a which reach the wirings 230 are formed in the planarized insulating film 250. Next, a pixel electrode 270 connected to the source/drain 7 via the wirings 230 is formed on the planarized insulating film 250. The pixel electrode 270 is formed as a transparent electrode or a reflection electrode depending on the display type of the liquid crystal display device. FIG. 20A shows the cross section of the essential part of one pixel. Thereafter, though not illustrated, an alignment film covering the pixel electrode 270 is formed on the planarized insulating film 250, thus completing a drive substrate 290. Meanwhile, as shown in FIG. 20B, an opposing substrate 310 to be disposed opposite to the drive substrate 290 is prepared. The opposing substrate 310 has a common electrode 350 provided on a transparent substrate 330 and covered with an alignment film (not shown). The common electrode 350 is formed by a transparent electrode. With the drive substrate 290 and the opposing substrate 310 are disposed opposite to each other via a spacer 370 with the pixel electrode 270 and the common electrode 350 facing each other. Then, a liquid crystal phase LC is filled and sealed between the substrates 290 and 310 separated at a predetermined clearance from each other via the spacer 370, thus completing a liquid crystal display 410. In case of preparing an organic EL display device using the drive substrate 290 with the foregoing structure, the pixel electrode provided at the drive substrate 290 is set as a anode (or cathode), an organic layer having necessary functions, such as those of a hole injection layer, an emission layer and an electron transport layer, is laminated on the pixel electrode, and a common electrode is formed as a cathode (or anode) on the organic layer. Referring to FIGS. 16 and 17, the thin film semiconductor device 100 acquired by using the crystallization method according to the embodiment described above is configured to have the grain boundaries a extending along the gate electrode 5, crossing the channel portion C, and periodically arranged in the direction of the channel length. Accordingly, the carriers which pass through the channel portion C always move crossing the grain boundaries a arranged at the predetermined width W1 therebetween. It is therefore possible to accurately control the transistor characteristic (carrier mobility) of the thin film transistor TFT in the thin film semiconductor device 100 by controlling the width (i.e., the width W1 of the belt-like crystal grains b). In other words, a variation in carrier mobility in a plurality of devices is suppressed by making the width W1 of the belt-like crystal grains b and the number of the belt-like crystal grains b arranged in the channel portion C (i.e., the number of the grain boundaries a) in one device equal to those of another device. Moreover, the crystal state between grain boundaries a and a is provided by the same crystal grains b. Accordingly, an amorphous region is not included, thus suppressing degradation of the device characteristic and keeping the carrier mobility in the direction of the channel length L high. It is therefore possible to prevent non-uniformity in luminance and non-uniformity in color at the display part by configurating a display device using the thin film transistors TFT formed in such a thin film semiconductor device as switching elements. The foregoing description of the embodiment has been given of the method of preparing a thin film semiconductor device having thin film transistors by adopting the semiconductor processing method according to the invention. The semiconductor processing method according to the invention is not however limited to the application to a method of fabricating a thin film transistor, and can be adapted to methods of fabricating other electronic devices. In any case, an electronic device with a good characteristic accuracy can be acquired by setting the current to flow in a direction crossing the grain boundaries a. Further, the materials, sources, processes, numerical values, etc. illustrated in the foregoing description of the embodiment are to be considered as illustrative, and materials, sources, processes, and numerical values different from the illustrated ones may be used as needed. The present application contains subject matter related to that disclosed in Japanese Priority Patent Application JP filed in the Japan Patent Office on Jul. 30, 2008, the entire contents of which is hereby incorporated by reference. It should be understood by those skilled in the art that various modifications, combinations, sub-combinations and alterations may occur depending on design requirements and other factors insofar as they are within the scope of the appended claims or the equivalents thereof. |
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044477309 | description | DETAILED DESCRIPTION OF THE INVENTION Referring now to the drawings, a transportation and/or storage container embodying this invention comprises mainly a thick-wall container body 12 of metallic material for containing therein and shielding gamma radiation of radioactive material. Such a container is described in the copending application of Botzem et al, Ser. No. 282,015 filed July 10, 1981, and assigned to the same assignee as this application. The disclosure of that application is incorporated by reference herein. The material of the body 12 desirably is made of cast iron, preferably spherical graphite cast iron. The container body 12 preferably is cylindrical and is provided with a plurality of preferably equally-spaced exterior cooling ribs 62 which may be arranged longitudinally, as shown, or circumferentially. These ribs 62 may be either cast integrally with the body 12 or welded thereto. Extending transversely to and between the ribs 62 and secured to the latter and to the exterior of the container body 12 are bridges or flanges 64 of a height less than that of the cooling ribs. These flanges 64 preferably are cast integrally with the body 12 and ribs 62, or they may be welded in place. The presence of the flanges 64 insures that in the event of the breaking off or cracking off of a cooling rib 62, the line of fracture of such rib is not located at its base but at a distance above such base. This minimizes the possibility that the breaking off or cracking off of a rib 62, which may occur in the event the container is dropped, will extend the fracture into the container body 12, i.e., insures against the continuation of such a crack into the container body. The safe distance of such a rib crack from the container body 12 itself can be insured by the judicious selection of relative heights of the flanges 64 and the ribs 62. Theoretical calculations, and also experiments, have shown that it is desirable to proportion the parts such that the height of the flanges 64 is no more than about two-thirds of the height of the cooling ribs 62 and that the spacing between flanges is no more than about ten times their height. In connection with the foregoing provision for inhibiting the extension of rib fracture cracks into the body 12 of the container, it has been found to be advantageous to provide notches 66 in the cooling ribs 62 in the area of intersection therewith of the flanges 64, as shown in FIG. 4. This construction even better insures against continuation of a rib crack into the body 12 of the container. The bottom of the notches 66 is no deeper than the tops of the flanges 64. Notches 66 of lesser depths are effective for their intended purpose, however, depending upon the specific design and material of the ribs 62 and the flanges 64. Such notches 66 may be molded by a casting operation or formed by a machining operation. In a working example, a cast body container for irradiated nuclear fuel elements taken from a pressurized water reactor, the container having cast on longitudinal cooling ribs and circumferential flanges, may have a flange spacing of the order of about 440 mm and a flange height of about 70 mm. The cooling ribs may have a height of about 240 mm with notches therein, at the location of intersection of the flanges, of a depth of the order of about 95 mm. With these dimensions, potential cracks in the container body 12, occasioned by damage to the cooling ribs, will be avoided. Neutron shielding material (not shown) desirably covers the cylindrical outer surface of the container body 12, preferably being disposed between the cooling ribs 62 in the spaces between the flanges 64 and the ribs. |
abstract | A diagnostic formulation is provided comprising a tropane having a radioactive concentration of at least 1.6 mCi/mL at least about 51 hours post creation. The diagnostic formulation optionally comprises a radiolabeled dopamine transporter (DAT) ligand useful in the diagnosis of Parkinson's disease (PS). One example of a radiolabeled dopamine transporter (DAT) ligand example is [123I]-2β-carbomethoxy-3β-(4-flurophenyl)-N-(3-iodo-E-allyl) nortropane. |
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claims | 1. A method of axially repositioning an annulus spacer having an inner cylindrical surface in contact with an inner tube and an outer cylindrical surface in contact with a generally coaxial outer tube, which method comprises the steps of:(a) vibrationally isolating a section of the wall of the inner tube adjacent to the annulus spacer;(b) causing said annulus spacer to go from a loaded condition to an unloaded condition;(c) vibrating the annulus spacer by vibrating the isolated section of the wall at a desired frequency such that the annulus spacer is displaced longitudinally from an initial position to a required position, whereby the vibration of the annulus spacer produces accelerations sufficient to overcome the tension of the annulus spacer on the inner tube. 2. The method of claim 1, wherein said step of vibrationally isolating the section of the inner tube comprises applying pressure to the inner surface of the wall of the inner tube at a first and a second position, which first and second position define the boundaries of said vibrationally isolated section. 3. The method of claim 1, wherein in step (c) the vibration of the annulus spacer is continued for a predetermined length of time. 4. The method of claim 1, wherein prior to step (a), the position of the annulus spacer is determined by:(i) vibrationally isolating a section of the inner tube;(ii) vibrating the wall of said inner tube within said isolated section;(iii) detecting vibration of the wall at a minimum of two axial positions within said isolated section; and(iv) detecting a reduction in vibration level of the wall at one or more of said axial positions in comparison to the remaining axial position(s),wherein the reduction in vibration level is indicative of the presence of the annulus spacer at or near the axial position at which said reduction in vibration was detected. 5. The method of claim 4, wherein said step of vibrationally isolating the section of the inner tube comprises applying pressure to the inner surface of the wall of the inner tube at a first and a second position, which first and second position define the boundaries of said vibrationally isolated section. 6. The method of claim 4, wherein following step (c) the new position of the annulus spacer is determined by:(v) vibrationally isolating a section of the inner tube;(vi) vibrating the wall of said inner tube within said isolated section;(vii) detecting vibration of the wall at a minimum of two axial positions within said isolated section; and(viii) detecting a reduction in vibration level of the wall at one or more of said axial positions in comparison to the remaining axial position(s),wherein the reduction in vibration level is indicative of the presence of the annulus spacer at or near the axial position at which said reduction in vibration was detected and wherein steps (a)-(c) are repeated as necessary to move the annulus spacer to a prescribed location about said inner tube. 7. The method of claim 6, wherein said step of vibrationally isolating the section of the inner tube comprises applying pressure to the inner surface of the wall of the inner tube at a first and a second position, which first and second position define the boundaries of said vibrationally isolated section. 8. The method of claim 4, wherein in step (c) the vibration of the annulus spacer is continued for a predetermined length of time. 9. The method of claim 1, wherein following step (c) the new position of the annulus spacer is determined by:(v) vibrationally isolating a section of the inner tube;(vi) vibrating the wall of said inner tube within said isolated section;(vii) detecting vibration of the wall at a minimum of two axial positions within said isolated section; and(viii) detecting a reduction in vibration level of the wall at one or more of said axial positions in comparison to the remaining axial position(s),wherein the reduction in vibration level is indicative of the presence of the annulus spacer at or near the axial position at which said reduction in vibration was detected and wherein steps (a)-(c) are repeated as necessary to move the annulus spacer to a prescribed location about said inner tube. 10. The method of claim 9, wherein said step of vibrationally isolating the section of the inner tube comprises applying pressure to the inner surface of the wall of the inner tube at a first and a second position, which first and second position define the boundaries of said vibrationally isolated section. 11. The method of claim 1, wherein said inner tube is a pressure tube and said outer tube is a calandria tube of a CANDU reactor fuel channel. |
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050080695 | summary | BACKGROUND AND SUMMARY OF THE INVENTION The present invention relates to a device for cooling of a heat-generating member which is placed in a liquid located in a pressure vessel, which liquid is adapted to serve as a coolant for the heat-generating member. The present invention is specially directed to the case where the heat-generating member consists of a core in a nuclear reactor of the type in which the reactor vessel is arranged in a neutron-absorbing liquid, for example an aqueous solution of boric acid, and the reactor coolant in the lower and upper ends of the reactor is arranged in hydraulic communication with the neutron-absorbing liquid. During normal operation the core in such a reactor is cooled by means of a coolant flow, normally a light water flow, continuously flowing through the core and the reactor vessel, the heat contents of the flow being utilized in a steam generator, or other heat exchanger, before the coolant is returned to the reactor vessel for repeated heating in the core. A reactor of such a kind is described, inter alia, in Swedish patent specifications Nos. 7606622-4 and 8102000-0. In case of an operational disturbance, an established pressure equilibrium between the neutron-absorbing liquid and the reactor coolant is disrupted, resulting in the neutron-absorbing liquid flowing into the reactor core and in the rest of the reactor, so that the power thereof is reduced to the decay power. The core with its "inherent" heat generation is then situated in the neutron-absorbing liquid. Operating situations with decay power cooling of the reactor core may occur both with a pressurized pressure vessel and with the vessel pressure-relieved to a pressure which only somewhat exceeds the atmospheric pressure. In certain hypothetical cases the pressure may be almost as great outside the pressure vessel as inside the same. The reason for the pressure relief may be rupture on the pressure vessel or on a pressurized tube connected to the pressure vessel. Operating situations with decay power cooling of the reactor core may also occur, inter alia, in the event of interruption in the supply of electric power or in case of disturbance in the supply of coolant to the reactor core. In the case of decay power cooling of the core, the heat can be removed passively from the core with the aid of a medium flow which may consist of a liquid or steam or a combination thereof. The present invention makes it possible to ensure an efficient cooling of the liquid in the pressure vessel in nearly all operating situations with decay power cooling. The favourable result obtained according to the invention can primarily be attributed to the fact that the cooling of the liquid in the pressure vessel takes place under self-circulation of a flow of liquid and/or of steam of the liquid with any contents of uncondensable gas in a circulation system containing an evaporator arranged in an evaporation pool, and to the fact that a small part of the flow is discharged from the circulation system to a discharge vessel, in which a lower pressure prevails than in the evaporator. By this discharge of a small part of the flow, a major accumulation of uncondensable gas on the primary side of the evaporator is avoided, which results in a more efficient cooling of the flow in the circulation system upon passage through the evaporator. |
description | This application claims the benefit of U.S. Application Ser. No. 61/539,631, filed Sep. 27, 2011, the disclosure of which is incorporated herein by reference in its entirety. Positron Emission Tomography (PET) devices employ positron-emitting radionuclides which are typically introduced into a subject, such as a patient, in a pharmaceutical composition. The positrons emitted by the positron-emitting radionuclides collide with the subject under investigation, resulting in the emission of pairs of gamma rays, which are detected. PET imaging devices are widely used to diagnose cancer recurrences, metastases of cancer, whether an early stage of cancer is present or not, and, if cancer has spread, its response to treatment. PET is also used in diagnosing certain cardiovascular and neurological diseases by highlighting areas with increased, diminished, or no metabolic activity. Short-lived PET radionuclides suitable for use in PET devices include positron emitters having a half-life which is typically less than 5 days, and generally less than one day, such as Fluorine (F-18) (half-life 110 minutes), Carbon 11 (C-11) (half-life 20 minutes), Nitrogen 13 (N-13) (half-life 10 minutes), Oxygen-15 (O-15) (half-life 2 minutes), Iodine 124 (I-124) (half-life 4.2 days), Rubidium 82 (Rb-82) (half-life 75 seconds), Copper 64 (Cu-64) (half-life about 0.5 days), in quantities that are appropriate or required for dosing. Because of the short half-life of these radionuclides, they are unsuited to use in a calibration source for calibrating the PET device. Accordingly, PET calibration sources have been developed which include radionuclides which have a much longer half-life than the short-lived radionuclide used in imaging. These include radionuclides such as Germanium 68 (Ge-68) (half-life about 271 days) and Sodium 22 (Na-22) (half-life about 2.6 years). Methods have been developed to calibrate these long-lived radionuclides against the short-lived radionuclide. See, for example, U.S. Pat. No. 7,825,372 entitled SIMULATED DOSE CALIBRATOR SOURCE STANDARD FOR POSITRON EMISSION TOMOGRAPHY RADIONUCLIDES, and U.S. Pat. No. 7,615,740, issued Nov. 10, 2009, entitled SYRINGE-SHAPED DOSE CALIBRATION SOURCE STANDARD, both by Keith C. Allberg, the disclosures of which are incorporated herein by reference in their entireties. One problem with the use of such calibration sources is that PET devices differ by manufacturer and facilities such as hospitals, often have two or more different PET devices. Thus a single calibration source often cannot be used to calibrate the different PET devices. A facility thus often has keep two or more different calibration sources in stock. Additionally, it is difficult to compare the results of two different PET devices, since this would require cross calibrating the two calibration sources at the same time. There remains a need for a system and method for enabling a calibration source to be used interchangeably in two or more PET devices. Aspects disclosed relate to a universal mounting adapter, an assembly including the adapter, a method of making the adapter and assembly, a calibrated source that can be used on the different PET devices and a method of use of the assembly. The adapter is configured for removable interconnection with two imaging devices allowing both to be calibrated with the same calibration source in the prescribed geometry where the two imaging devices are incompatible in terms of their ability to mount a conventional calibration source. In accordance with one aspect of the exemplary embodiment, an assembly includes a calibration source which includes a radionuclide; and an adapter connected to the calibration source. The adapter includes a first mounting mechanism adapted for mounting the adapter to a first mounting bracket of a first imaging device whereby the calibration source is positioned for calibrating the first imaging device. The adapter also includes a second mounting mechanism adapted for mounting the adapter to a second mounting bracket of a second imaging device, the second mounting bracket being different from the first mounting bracket, whereby the calibration source is positioned for calibrating the second imaging device. In accordance with another aspect of the exemplary embodiment, a universal mounting adapter is provided for mounting an associated calibration source in associated first and second imaging devices. The adapter includes a plate including first and second opposed planar surfaces and a peripheral surface which connects the planar surfaces. A threaded shaft extends from a center of the first surface of the plate. Two studs extend from the second surface of the plate. An arcuate slot is defined in the peripheral surface which extends around at least a portion of the peripheral surface. In accordance with another aspect of the exemplary embodiment, a method for calibrating two imaging devices is provided. The first imaging device includes a first mounting bracket and the second imaging device including a second mounting bracket different from the first mounting bracket. The method includes providing a calibration source which includes a radionuclide and mounting an adapter to the calibration source. The adapter includes a first mounting mechanism adapted for mounting the adapter to the first mounting bracket and a second mounting mechanism adapted for mounting the adapter to the second mounting bracket. The method further includes mounting the adapter to the first mounting bracket of the first imaging device using the first mounting mechanism but not the second mounting mechanism, whereby the calibration source is positioned for calibrating the first imaging device and, thereafter, mounting the adapter to the second mounting bracket of the second imaging device using the second mounting mechanism but not the first mounting mechanism, whereby the calibration source is positioned for calibrating the second imaging device. In accordance with another aspect of the exemplary embodiment, a method of making an assembly for calibrating two imaging devices is provided. The first imaging device includes a first mounting bracket and the second imaging device includes a second mounting bracket different from the first mounting bracket. The method includes providing a calibration source which includes a container which holds a radionuclide, the container including a threaded bore in an end wall and mounting an adapter to the calibration source, the adapter comprising a first mounting mechanism adapted for mounting the adapter to the first mounting bracket and a second mounting mechanism adapted for mounting the adapter to the second mounting bracket and a threaded shaft which is received within the threaded bore. With reference to FIG. 1, an assembly 10 comprising a calibration source 12 and a universal mounting adapter 14 adapted for selectively mounting the calibration source 12 to a mounting bracket of an imaging device according to the exemplary embodiment is illustrated. The calibration source 12 is designed to provide a calibrated radiation dose when positioned in the imaging device. The exemplary imaging device is one which detects positrons, such as a PET imaging device or a device which combines PET with one or more other imaging methods, such as PET/CT or the like. In the case of PET imaging, the calibration source 12 may contain a calibrated quantity of Na-22 or Ge-68/Ga-68 with a determined/determinable F-18 equivalent value. The calibration source 12 (FIG. 2) includes a container 16 which includes a cylindrical barrel 18 and a closure member 20 mounted to a first end of the barrel, which seals a radioactive dose 21 within container 16. The barrel 18 includes a cylindrical side wall 22 of substantially uniform cross section which is closed at a second end by an end wall 24. The end wall 24 may be integrally formed with the side wall 22, for example by machining from a single piece of plastic, molding or otherwise fabricating from a single piece. The container may be formed from a rigid plastic, such as high density polyethylene (HDPE). The closure 20 may be attached to the barrel by screws 28 (FIG. 2) and/or a sealant, or other fastener member(s). Screws 28 can be formed of nylon, for example. The mounting adapter 14 (FIGS. 3-5) includes a generally circular plate 30 with first and second opposed planar surfaces 32, 34 spaced by a peripheral surface 36. An externally threaded shaft 38 extends from the plate 30 in a direction perpendicular to the planar surface 34. The plate and threaded shaft may be integrally formed e.g., by machining them from a single piece, molding, or the like. The adapter 14 may be formed, for example from aluminum (e.g., at least 50% by weight aluminum, or at least 70 wt. % or at least 90 wt. %, or about 95 wt. % aluminum), such as an aluminum alloy, or other metal or other material which is rigid and ideally resistant to wear and corrosion. The material used for forming the adapter may have a yield strength of at least 140 MPa, e.g., at least 200 MPa, a tensile strength of at least 200 MPa, e.g., at least 250 MPa, and an elongation at break of less than 10%. As an example, a precipitation hardening aluminum alloy, containing magnesium and silicon as its major alloying elements can be used, such as a 6061 alloy, e.g., a tempered alloy, such 6061-T6 aluminum alloy (solutionized and artificially aged) is used. As shown in FIG. 2, the container end wall 24 is of sufficient thickness to accommodate a threaded bore 40 adapted for threadably receiving and engaging the threaded shaft 38 of the mounting adapter 14. Both the bore 40 and the threaded shaft 38 may be double or triple threaded with complementary threads for creating a rigid engagement with virtually no play. A planar exterior surface 42 of the end wall 24 contacts the surface 34 of the plate when the threaded shaft and bore are fully engaged. The container 16 defines a cylindrically-shaped interior cavity 46 which holds the radioactive source-containing material 21, sealed within the barrel 18 by the planar closure member 20. The exemplary radioactive source-containing material 21 may include one or more radionuclides encapsulated in a suitable solid matrix material. Exemplary nuclides include gamma radiation emitters, such as germanium 68 (Ge-68) or sodium 22 (Na-22), in appropriate quantities for serving as a traceable calibration source that acts as a proxy for F18. The matrix material may comprise an epoxy, silicone, urethane, ceramic, or similar type of matrix material in which the radionuclide may be uniformly dispersed to form a solid mixture. For example, the calibration source 12 may include radioactive material having an activity of from 0.1-20 millicuries (mCi). While FIG. 1 shows the interior being entirely filled with radioactive source-containing material 21, there may be an air space between the material and the closure 20. The exemplary plate 30 has a diameter D which is greater than a diameter d of the container 12, such that the plate overhangs the container, as seen in FIG. 1. The barrel 18 has an interior length L which sufficient to present a suitable length of radioactive material 21 to the imaging device for positrons emitted when the gamma radiation collides with container to be detected by the detectors of the imaging device. The exact length may be dependent on the type of imaging devices in which it is to be used, e.g., whether the imaging devices are one, two, or three ring devices. L may be, for example, from about 2-100 cm, such as about 20-40 cm. In one embodiment, the outer barrel diameter d may be about 2-50 cm e.g., about 6-20 cm, and plate diameter D about 0.5-10 cm greater than d, e.g., the barrel diameter d may be about 20 cm and the plate diameter D may be about 25 cm in diameter. The volume of the interior cavity 46 may be from about 3 to about 20,000 cm3, such as at least about 500 cm3, e.g., about 7000 cm3. Referring once more to FIGS. 3 and 4, the mounting adapter 14 provides plural types of mounting mechanisms 50, 52 for selectively mounting the adapter 14, and hence the calibration source 12, to suitably configured brackets of different imaging devices. In particular, a first mounting mechanism 50 is in the form of a single groove or slot which extends, from the peripheral surface 36, into the plate 30. The slot 50 extends parallel to and intermediate the surfaces 32, 34. As best seen in FIG. 5, the slot 50 is arcuate, e.g. annular in shape. The exemplary slot has an inner radius r of about 8-10 cm, e.g., r=d/2 and a uniform, radial width w (extending from the peripheral surface 36 of the plate 30 to the inner radius r) which may be less than ¼D, such as about 1-5 cm, e.g., 2-4 cm. The slot 50 extends at least partially around the circumference of the plate 30, to subtend an angle ⊖ of at least 45 degrees or at least 90 degrees or at least 120 degrees, or at least 160 degrees, e.g., about half way round the plate (e.g., ⊖=160-190°, or ⊖<180°). The slot has a depth f (perpendicular to the surfaces of the plate) of about 0.5 to 1 cm (FIG. 3) which is sized to receive a portion of a bracket 54 of the imaging device therein (FIG. 6). In particular, the bracket is in the shape of a rectangular plate with an arcuate (e.g., semi-circular) cut out 56 of approximately the same radius r as the slot's inner radius. In this way, the mounting adapter can be slotted into the slot, holding the assembly 10 rigidly positioned with respect to x, y and z axes (FIG. 9). The bracket has a width n which is only slightly less than the width f of the arcuate slot 50, so that it is received within the slot 50 up to a depth of about w, and firmly gripped by the side walls of the slot. The bracket 54 is thus designed to position a longitudinal axis X of the calibration source 12 along the central axis x of the imaging device, i.e., in the prescribed geometry for the calibration source 12 to calibrate the imaging device. It is not necessary to prevent rotation of the assembly about the x axis since the barrel is symmetrical about the X axis. As will be appreciated, although not shown, the bracket 54 may include additional members, e.g., one in contact with each of surfaces 32 and 34 of the plate, to provide additional support for the assembly. Engagement of the adapter with the bracket 54 in the prescribed geometry, and subsequent disengagement from the bracket, can be performed entirely without tools, i.e., by hand. As shown in FIG. 5, the second mounting mechanism 52 includes a pair of studs 53A, 53B which extend from the surface 32 of the plate. The studs each include a generally cylindrical shank 58, of length g and diameter j, and an enlarged head 60 at a terminal end of the shank (FIG. 4). The head has a diameter k, where k>j. The studs 53A, 53B may be integrally formed with the plate 30 to provide a shank with a fixed length. Alternatively, the studs may be fitted with a threaded end 62 (FIG. 7) for fastening the stud to the plate and optionally for variably adjusting the exposed length g of the shank 58. The studs 53A, 53B are equidistant from the shaft 38 and may be spaced apart by a distance h (FIG. 5) of approximately 2×r such that the center of the studs lie on the same radius as an inner end of the slot. In one embodiment, the centers of the studs 53A, 53B and the shaft 38 are collinear. The studs 53A, 53B are configured for mounting the assembly 10 to a second mounting bracket 64 which includes a pair of slots 66A, 66B of uniform width, defined in an upper end thereof (FIG. 8). The bracket slots 66A, 66B are spaced from each other by a distance h to receive a respective stud shank 58 therein. The slots 66A, 66B are open at each side of the bracket 64 to allow the studs 53A, 53B to extend therethrough. As will be appreciated, while two slots 66A, 66B and two studs 53A, 53B are shown, more than two of each could be employed. The bracket 64 can be in the form of a rectangular or otherwise shaped plate with a thickness that is the same as the shaft length g so that the head of the stud and surface 32 of the circular plate grip either side of the bracket 64 tightly when the assembly 10 has been slid into place (FIG. 10). The slots in the bracket have a length M which is selected to position the longitudinal axis X of the source along the central axis x of the imaging device (FIG. 10), i.e., in the prescribed geometry for the imaging device. Engagement of the adapter with the bracket 64 in the prescribed geometry, and subsequent disengagement from the bracket, can be performed entirely without tools, i.e., by hand. As shown in FIG. 9, the first bracket 54 is mounted to a patient table 70 of a first imaging device 71. The table is designed to support a subject, such as a person or animal, during an imaging procedure. The patient table 70 moves in the x direction through a ring of detectors 72 arranged in pairs offset by 180° (only two are shown for ease of illustration). The detectors generate electrical signals in response to the detection of positrons, which are processed by a detection system 74 to generate a PET image of the subject, who has been dosed with a short lived radionuclide, such as F18. During calibration, the detectors 72 provide calibration signals, in response to detection of positrons emitted from the calibration source 12, which are used by the detection system 74 to provide a calibration for the short-lived radionuclide-based signals. As will be appreciated, the second bracket 64 is similarly rigidly mounted to a second patient table 76 (FIG. 10) of another imaging device with a ring of detectors and a detection system (not shown), which can be configured similarly to that shown in FIG. 9. The mounting mechanisms 50, 52 are arranged on the plate 30 so that irrespective of which of the two imaging devices the assembly is used in, the calibration source 12 is properly aligned with the respective ring of detectors. This allows the two imaging devices to be calibrated with the same calibration source, by moving the assembly 10 from one imaging device to the other. This allows reproducibility in calibration of the two devices and allows imaging results output by the two devices to be compared with greater accuracy. The calibration source 12 may be marked with suitable markings 80 on the barrel which allow its position to be detected, e.g., with a laser, and any errors in its position corrected by adjustments to the respective mounting bracket 54 or 64. To form the assembly 10, a container 16 is formed by machining one end of a cylindrical a solid block of plastic to define the interior cavity 46 and machining the other end to define the threaded bore 40. Appropriate quantities of a radionuclide (e.g., Ge 68) in liquid form and a liquid polymer composition are mixed to disperse the radionuclide uniformly in the polymer (having saved some of the radionuclide liquid or liquid mixture for testing to be calibrated e.g., against a traceable National Institute of Standards (NIST) solution of F18, as described, for example, in above-mentioned U.S. Pat. No. 7,825,372). The polymer composition may include a polymer resin together with accelerators, crosslinking agents, and the like which cause the polymer to harden when cured (e.g., by UV-curing or an ambient cure). The liquid radionuclide/polymer composition is placed in the barrel 18 and cured to form a solid 21. The barrel is then sealed to the closure member 20, for example, by placing a small amount of the polymer matrix material around the end of the barrel and then screwing the screws 28 into the barrel. A custom decay calendar may then be derived and a label affixed to the calibration source or to a shielding container in which the source 12 is shipped and stored. The exemplary label also carries the conversion factor(s) for one or more PET radionuclides, such as F18. The completed cylinder source 12 can then be stored and/or shipped, e.g., in a radiation shielded case. The adapter 14 can be affixed to the cylinder source 12 at any suitable time, and optionally removed therefrom after use. To cross calibrate two imaging devices, the assembly 10 is mounted to a first of the imaging device brackets (e.g., bracket 54) and the table advanced through the ring of detectors while signals generated thereby are received at the detection system 74 and processed. The assembly is removed from the first mounting bracket and mounted to the second mounting bracket 64 and the calibration process is repeated. By comparing the results of the two scans, any differences between the two imaging devices can be minimized by modifying the algorithm which converts the signals received from imaging a subject to a resulting image. It will be appreciated that variants of the above-disclosed and other features and functions, or alternatives thereof, may be combined into many other different systems or applications. Various presently unforeseen or unanticipated alternatives, modifications, variations or improvements therein may be subsequently made by those skilled in the art which are also intended to be encompassed by the following claims. |
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claims | 1. A device comprising:a luminescent material, the luminescent material being a parallelepiped having a first side and a second side that are squares and having a third side that is a rectangle or a square, the first side and the second side being perpendicular to the third side;a beam stop plate attached to the first side of the luminescent material; andan optical fiber having a first end and a second end, the first end of the optical fiber attached to the third side of the luminescent material. 2. The device of claim 1, wherein the luminescent material is selected from a group consisting of cerium-doped yttrium aluminum garnet, lead tungstate, cadmium tungstate, and diamond. 3. The device of claim 1, wherein first side and the second side of the luminescent material are about 250 microns to 1 millimeter by about 250 microns to 1 millimeter, and wherein the third side of the luminescent material is about 250 microns to 1 millimeter by about 250 microns to 1 millimeter. 4. The device of claim 1, wherein the luminescent material comprises three or four blocks of material, and wherein each block of material has dimensions of about 250 microns to 1 millimeter by about 250 microns to 1 millimeter by about 80 microns to 350 microns. 5. The device of claim 1, wherein the optical fiber comprises a multi-mode optical fiber. 6. The device of claim 1, wherein the optical fiber has a circular cross section having a diameter of about 350 microns to 450 microns. 7. The device of claim 1, wherein the optical fiber has a length of about 10 centimeters to 2 meters. 8. The device of claim 1, wherein the beam stop plate comprises a metal selected from a group consisting of silver, gold, tantalum, and tungsten. 9. The device of claim 1, wherein the beam stop plate has circular shape and a diameter of about 250 microns to 2 millimeters. 10. The device of claim 1, wherein the beam stop plate has an oval shape with an about 250 micron to 2 millimeter short axis and an about 300 micron to 3 millimeter long axis. 11. The device of claim 1, wherein the beam stop plate is about 100 microns to 1 millimeter thick. 12. The device of claim 1, further comprising:a hollow sleeve, wherein an end of the hollow sleeve is attached to the beam stop plate, and wherein the hollow sleeve includes a cutout to allow the optical fiber to contact the luminescent material. 13. The device of claim 12, wherein the hollow sleeve comprises a metal selected from a group consisting of silver, gold, tungsten, and tantalum. 14. The device of claim 12, wherein a wall of the hollow sleeve is about 100 microns to 300 microns thick. 15. The device of claim 12, wherein the beam stop plate has a circular shape, wherein the hollow sleeve is a hollow cylinder, and wherein an outer diameter of the hollow sleeve is the same as a diameter of the beam stop plate. 16. The device of claim 1, further comprising:a photodiode, wherein the photodiode is attached to the second end of the optical fiber. 17. The device of claim 1, wherein the optical fiber is one of a plurality of optical fibers that form a bundle of optical fibers. 18. The device of claim 17, wherein each optical fiber of the bundle of optical fibers has a circular cross section having a diameter of about 100 microns to 300 microns. 19. A device comprising:a luminescent material, the luminescent material being a parallelepiped having a first side and a second side that are squares and having a third side that is a rectangle or a square, the first side and the second side being perpendicular to the third side;a beam stop plate attached to the first side of the luminescent material; anda bundle of optical fibers having a first end and a second end, the first end of the bundle of optical fibers attached to the third side of the luminescent material. 20. A method comprising:(a) providing a luminescent material, the luminescent material being a parallelepiped having a first side and a second side that are squares and having a third side that is a rectangle or a square, the first side and the second side being perpendicular to the third side;(b) attaching a beam stop plate to the first side of the luminescent material; and(c) attaching a first end of an optical fiber to the third side of the luminescent material. |
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052951707 | summary | BACKGROUND OF THE INVENTION This invention relates to techniques for adjusting the pH of post accident containment water in nuclear power plants, and in particular passive techniques. Nuclear power plants are generally provided with reactor containment vessels that surround the nuclear reactor core and steam generators attached to the core. Water, either pressurized or boiling, circulates through the core, extracting heat from the core for the generation of power. Following an accident that involves the loss of reactor coolant, an emergency core cooling system (ECCS) fills the reactor core with water to a level such that the core is covered. It is possible for the core to suffer some damage in such an accident, and, in the case where the safety injection system fails, to suffer significant damage. After an accident, radioactive iodine present in the containment vessel can change from a particulate form, such as CeI salts, to a more volatile organic form. The organic form of iodine behaves like a gas and is more likely to leak out of the containment vessel, thereby resulting in greater releases of radioactivity from the plant. Adjusting the pH of the water in the containment vessel prevents the radioactive iodine from changing into the organic form. Therefore, during and after such accidents, it is necessary to adjust the pH of the water that is in the containment sump in order to limit the amount of radiation released from the plant. Initially after an accident, the pH of the water in the containment vessel may be as low as 3. In order to prevent the formation of significant amounts of organic iodine the pH must be raised to a value of about 7. In current plants, the pH of the post accident water is adjusted by the addition of chemicals such as sodium hydroxide. This is typically accomplished by the use of containment spray pumps to inject sodium hydroxide along with the spray water from the ECCS. This provides an effective means of adjusting the pH. However, this means of injection is subject to some of the same failure mechanisms that can cause the failure of the safety injection system. Some of the common cause failure mechanisms include the complete loss of all AC power, the complete loss of all cooling water, or the complete loss of all ventilation systems. In some recent advanced nuclear reactors, pumps are not used in the safety systems, and in some passive plants there is no spray system at all. Therefore, a different means of adjusting the pH is required. One approach is to place water permeable sacks of NaOH crystals in the ECCS sump. When the sump is flooded, the NaOH dissolves in the water. However, this passive system suffers serious drawbacks. First, the NaOH is very caustic and must be isolated from equipment. Second, accidental water spills in the containment vessel could activate this type of passive system, necessitating expensive clean-up procedures to remove the NaOH and to repair any damaged equipment. Third, the sacks must be checked for leaks and replaced frequently. Therefore, it would be advantageous to provide a passive pH adjustment system that overcomes the above stated problems. SUMMARY OF THE INVENTION An object of this invention to provide a method of adjusting the pH of post accident water in a nuclear reactor containment vessel. An additional object of this invention is to provide a nuclear reactor having a means of adjusting the pH of post accident water in a nuclear reactor containment vessel that does not depend upon the safety injection system. It is a further object of this invention to provide a passive means of adjusting the pH of post accident water in the nuclear reactor containment vessel that will be activated only in an emergency situation. According the invention, a nuclear reactor is placed within a containment vessel. Means are provided for elevating the pH of post accident water in the containment vessel. A tank containing a basic liquid, such as liquid sodium hydroxide, is located within the containment vessel at an elevation above the maximum post accident water level. A sensor provides a signal indicative of an elevated radiation condition within the containment vessel that is greater than a predetermined, normal operating radiation level. If the sensor detects a radiation condition much greater than the normal operating range of the nuclear reactor, it triggers an explosive squib valve attached to the bottom of the tank. This allows the basic liquid contained within the tank to drain by gravity into the containment vessel. The basic liquid is directed by routing means from the tank to two spaced apart pH sumps in the containment vessel. Each of the pH sumps is located proximate to a safety injection sump. The pH sumps are capable of holding the entire contents of the tank. A vacuum breaker is positioned near the top of the tank. Opening the vacuum breaker allows the basic liquid to more easily drain from the tank after the explosive valve has opened. |
description | [Embodiment 1] FIG. 1 is a schematic and front view of a vertical type stage system according to a first embodiment of the present invention. In this embodiment, the invention is applied to a vertical type X-Y stage which is incorporated into an X-ray exposure apparatus using synchrotron radiation, for supporting a wafer along a vertical direction and which is movable along the surface of the wafer. The X-Y stage system is provided with inertial force applying means and gravity compensating means. Denoted in FIG. 1 at 1 is a base for supporting the vertical type stage system, and denoted at 2 is an anti-vibration damper for removing or reducing vibration of the base 1. Denoted at 3 is a stage base which is supported by the base 1 and which has a reference surface for supporting the stage system. Denoted at 4 is a wafer chuck for holding a wafer thereon, and denoted at 5 is a main stage (second stage) for supporting the wafer chuck 4 and being movable along the wafer surface (X-Y plane). The main stage 5 has an X measurement mirror 6 and a Y measurement mirror 7 fixedly mounted at an end face thereof, for providing reflection surfaces for a position measurement beam. Denoted at 8 is a Y stage guide fixedly mounted on the stage base 3, and denoted at 9 is a Y stage base (first stage) being movable in the Y direction with the main stage 5 mounted thereon. The Y stage guide 8 serves to support the Y stage base 9 with respect to the X direction (second direction) and to guide it with respect to the Y direction (first direction) without contact thereto. Denoted at 10 is an X stage guide fixedly mounted on the Y stage base 9. It serves to support the main stage 5 with respect to the Y direction and to guide it with respect to the X direction without contact thereto. This embodiment uses an air bearing mechanism for non-contact guiding. However, any other mechanism may be used, as long as it can provide a function of low friction guiding. Denoted at 51 is a movable element of a linear motor (first driving mechanism) for providing a drive in the Y direction. It is fixedly mounted on the Y stage base 9, and is disposed opposed to a linear motor stator (not shown) mounted on the Y stage guide 8. The main stage 5 is provided with a movable element (not shown) of another linear motor (second driving mechanism) for providing a drive in the X direction. It is disposed opposed to an X linear motor stator 54, mounted on the X stage guide 10. Preferably, these linear motors are set so that their operational axis, along which the thrust is applied, extends through the gravity center of the element to be driven thereby. Denoted at 23 is a Y mass member (countermass) for applying an inertial force in the Y direction to the stage base 3. Denoted at 24 is a Y mass guide fixedly mounted on the stage base, for guiding the Y mass member 23 in the Y direction without contact thereto. As the Y mass member 23 moves in the Y direction, it functions as a Y-direction inertial force applying mechanism. Denoted at 27 is an X mass member (secondary countermass) for applying an inertial force in the X direction to the stage base 3. Denoted at 28 is an X mass guide fixedly mounted on the Y stage base 9, for guiding the X mass member 27 in the X direction without contact thereto. As the X mass member 27 moves in the X direction, it functions as an X-direction inertial force applying mechanism. Each of these masses is equipped with a movable element (not shown) of a linear motor for a drive in a predetermined direction, which is disposed opposed to the linear motor stator (not shown) provided on the corresponding guide, whereby an inertial force applying mechanism is provided. Preferably, these linear motors are set so that their operational axis, along which the thrust is applied, extends through the gravity center of the element to be driven thereby. This embodiment is provided with means for compensating for gravity applied to the Y stage, which comprises a cylinder mechanism with a piston, in this embodiment. Denoted in FIG. 1 at 31 is a cylinder rod (rod A) fixedly mounted on the Y stage base 9. An end thereof opposite to the Y stage base 9 is supported by a cylinder piston A32. Denoted at 33 is another cylinder rod (rod B) fixedly mounted on the Y mass member. An end thereof opposite to the Y mass member 23 is supported by a cylinder piston B34. The pistons A32 and B34 have a sealing function for a fluid within an air cylinder 35 of a coupled structure. Thus, through the cylinder mechanism 35, the weight of the Y stage base 9 as supported by the piston A32 and the weight of the Y mass member 23 as supported by the piston B34 are propagated, and they are balanced with each other. As a result of this, gravity compensation for the Y stage base 9 is accomplished. Also, the cylinder 35 as well as the pistons A32 and B34 are maintained in a non-contact state, for example, to assure movement with very low friction. Here, the weight of the Y stage base corresponds to total weight, including the main stage movable in the Y direction with the Y stage, the X stage guide, the X mass member, the X mass guide and the like. Similarly, the mass of the Y stage base corresponds to total mass, including an element movable in the Y direction with the Y stage base. Without mentioning otherwise, the same applies to other embodiments of the present invention to be descried later. The sectional area of the piston of the cylinder mechanism 35 is determined in consideration of the weight of Y stage base 9 and the weight of Y mass member 23. Details of this will be described later. A drive signal for the mass member is produced as follows. In the drawing, when the main stage 5 is to be driven in the X direction, the reaction force produced at the linear motor stator of the X stage guide as the main stage 5 is driven is denoted by FX, while the resultant force of a reaction force produced at the stator of the linear motor of the X mass guide 28 as the X mass member 27 of the inertial force applying mechanism is driven is denoted by RX. In order that the inertial force applying mechanism causes RX active to cancel FX, the following equation should be satisfied. RX=xe2x88x92FX Here, if the inertial force applying mechanism is arranged so that the operational axis of RX is registered with the operational axis of drive reaction force FX of the main stage 5, no rotational torque is produced at the main stage 5. Further, providing plural X mass members 27 will be effective to broaden the latitude for designing the structure so that the operational axis of the drive thrust applied to the main stage 5 extends through the gravity center position. However, use of plural X mass members 27 is not a requisition. Here, when the mass of the main stage is MX and the total mass of the X mass member 27 is mX, the driving stroke sX for the X mass member 27 can be determined, with respect to the X-direction stroke SX of the main stage, by the mass ratio between MX and mX. This can be expressed by an equation below. SX/sX=1/ (MX/mX) Namely, the ratio of movement amount of them displaces relative to the stage base 3, in a proportion corresponding to an inverse of the mass ratio. Thus, by making mX larger, the mass ratio MX/mX becomes smaller and, therefore, the driving stroke sX of the X mass member can be designed small. Since, however, the Y-direction movement mass including the X mass member 27 becomes larger, the energy required for Y-direction movement becomes larger. To the contrary, if sX can be designed large, the mass mX of the X mass member 27 can be made smaller and, therefore, the Y-direction movement mass can be made smaller. Therefore, the energy required for Y-direction movement is made smaller. The same applies to the Y direction. That is, as the Y stage base is driven in the Y direction, the Y mass member of the inertial force applying mechanism may be driven so as to cancel the reaction force produced at the Y stage guide 8. Since, however, there is a large influence of gravity, in regard to the Y direction, here, a description will be made only on the function of Y-direction driving means, similarly to the case of the X direction, on an assumption that a gravity compensating mechanism acting on the Y stage base 9 or Y mass member 23 is being operated such that the effect of gravity applied to the Y stage base 9 or the Y mass member 23 can be disregarded. Here, the resultant force of a reaction force produced at the Y stage guide 8 fixed to the stage base 3 as the Y stage base 9 is to be driven in the Y direction is denoted by FY, while the resultant force of a reaction force produced at the Y mass guide fixed to the stage base as the Y mass member 23 of the inertial force applying mechanism is driven is denoted by RY. In order that the inertial force applying mechanism causes RY active to cancel FY, the following equation should be satisfied. RY=xe2x88x92FY Here, if the inertial force applying mechanism is arranged so that the operational axis of RY is registered with the operational axis of drive reaction force FY, no rotational torque is produced at the stage base 3. Further, providing plural Y mass members 23 will be effective to broaden the latitude for designing the structure so that the operational axis of the drive thrust applied to the Y stage base extends through the gravity center position. However, use of plural Y mass members 23 is not a requisition. Here, when the mass of the Y stage base 9 is MY and the total mass of the Y mass member 23 is mY, the driving stroke sY for the mass member can be determined, with respect to the Y-direction stroke SY of the Y-direction movement element, by the mass ratio between MY and mY. This can be expressed by an equation below. SY/SY=1/(MY/mY) Namely, the ratio of a movement amount of them displaces relative to the stage base 3, in a proportion corresponding to an inverse of the mass ratio. Thus, by making mY larger, the mass ratio MY/mY becomes smaller and, therefore, the driving stroke sY of the Y mass member 23 can be designed small. Next, the balancing relation between Y stage base 9 and Y mass member 23 by the cylinder mechanism 35 as well as the relation between piston sectional area and piston movement amount, will be described. FIG. 2 shows a model of a cylinder mechanism which serves as a gravity compensating mechanism, in the first embodiment. In the cylinder mechanism 35, when the sectional areas of the pistons A32 and B34 are denoted by a and b, respectively, the relation between the total mass mY of the mass member 23 and the mass MY of the Y stage base 9 supported by the piston A32 can be expressed by an equation below. a/b =MY/mY When the Y stage base 9 or the Y mass member 23 is supported by plural pistons, the amount a or b corresponds to the total of the piston sectional areas. The ratio of movement amount between the pistons A32 and B34 whose sectional areas can be determined in accordance with the above equation, can be determined by an equation below, where SP is the movement amount of piston A32 and sP is the movement amount of piston B34. SP/sP=1(a/b) Namely, the sectional area ratio a/b of the piston A32 for supporting the weight of the Y stage base 9 and of the piston B34 for supporting the weight of the Y mass member 23, corresponds to an inverse of the movement amount ratio between the Y stage base 9 and the Y mass member 23. From these equations, it is seen that the movement amount ratio (SY/sY) of the Y mass member 23 and the Y stage base 9, constituting the stage system of this embodiment, and the movement amount ratio (SP/sP) of the pistons A32 and B34 of the piston mechanism (gravity compensating mechanism) are equal to each other. Because the movement amount ratio is the same, the Y mass member 23 and the Y stage base 9, constituting the inertial force applying mechanism can be connected to the gravity compensating mechanism through the cylinder rods A31 and B33. As a result, with a simple structure, both the inertial force applying mechanism and the gravity compensating mechanism are accomplished in parallel. Here, the ratio in sectional area of the pistons A32 and B34 is approximately equal to the mass ratio of the Y stage base 9 and the Y mass member 23. While in this embodiment air is used as the medium within the cylinder, this is not a requisition. A hydraulic cylinder, for example, may be used, as long as a similar function is provided. In this embodiment, a drive reaction force due to drive of the main stage in the X direction is reduced by movement of the X mass member and, as a result, application of vibration to be applied to the base or the stage base can be reduced. Additionally, since the main stage and the X mass member move in opposite directions with respect to the X direction, at a proportion corresponding to an inverse of the mass ratio, advantageously, there occurs substantially no shift of a gravity center position of the whole stage system. Therefore, deformation of the base which supports the stage base can be suppressed. Particularly, in a vertical type stage such as of the present embodiment, suppression of a shift of gravity center position or of a change in reaction force with respect to the X direction, is very effective to a high speed and high precision stage system, since it reduces a change in load applied to an anti-vibration damper. Further, in this embodiment, reaction force resulting from drive of the Y stage base in the Y direction is reduced by movement of the Y mass member and, as a result, application of vibration to be applied to the base or the stage base can be reduced. Additionally, since the Y stage base and the Y mass member move in opposite directions with respect to the Y direction, at a proportion corresponding to an inverse of the mass ratio, advantageously, there occurs substantially no shift of the gravity center position of the whole stage system. Therefore, deformation of the base which supports the stage base can be suppressed. Furthermore, since the weight of the Y stage base is compensated for by a cylinder mechanism of a coupled structure, both the inertial force applying mechanism and the gravity compensating mechanism are accomplished in parallel, with a simple structure. Thus, stage driving can be performed with less energy and less heat generation. [Embodiment 2] FIG. 3 is a schematic and front view of a vertical type stage system according to a second embodiment of the present invention. The components corresponding to those of the first embodiment of FIG. 1 are denoted by the same reference numerals. Like the first embodiment, the second embodiment is applied to a vertical type X-Y stage system which is equipped with an inertial force applying mechanism and a gravity compensating mechanism. The principle of operation and the manner of operation of the main stage or inertial force applying mechanism of the second embodiment are essentially the same as those of the first embodiment, and a description therefor will be omitted. In the second embodiment, gravity compensation is provided by a pulley and belt mechanism, in place of the gravity compensation mechanism having a cylinder mechanism. Denoted in FIG. 3 at 41 is a belt from which a Y mass member 23 and a Y stage 9 are suspended. Denoted at 42 is a fixed table which is fixedly mounted on a stage base 3. Denoted at 43 is a pulley unit disposed on the fixed table 42. As seen from the drawing, since the belt 41 is wound around the pulley unit 43, the pulley unit 43 functions to support the Y stage base 9 and the Y mass member 23 through the belt 41. FIG. 4 shows a model of a pulley unit which is a gravity compensating mechanism, in the second embodiment. Since the pulley unit 43 comprises pulleys each having two diameters, use of two belts is necessary for each pulley. An end of the belt 41 is fixed to the pulley, and an end of the other belt is attached to the Y stage base 9 or Y mass member 23. In the second embodiment, particularly in regard to the ratio of pulley diameters of the pulley unit 43, the balance of weight between the Y stage base 9 and the Y mass member 23 as well as the relation between the movement amount of Y stage base 9 and Y mass member 23 and the pulley diameter, will be described. Here, the mass of the Y stage base 9 is MY and total mass of the Y mass member 23 is mY. In the pulley unit 43, the diameter of the pulley on which the belt 41 coupled to the Y stage base 9 is wound is d, and the diameter of the pulley on which the belt 41 coupled to the Y mass member 23 is wound is e. In this case, for balancing the weight of Y stage base 9 and Y mass member 23 with the pulley unit 43, the pulley diameter ratio d/e of the pulley unit 43 is determined in accordance with the mass ratio MY/mY, and it can be expressed by an equation below. d/e=1/(MY/mY) Namely, the ratio of pulley diameters of the pulley unit 43 corresponds to an inverse of the mass ratio between the Y mass member 23 and the Y stage base as supported by the pulleys. Here, the movement amount SP of the Y stage base 9 and the movement amount sP of the Y mass member 23 as the pulleys are arranged with this pulley diameter ratio, are proportional to the diameters of the pulleys to which they are connected, respectively. Thus, they are expressed by an equation below. SP/sP=d/e From these equations, it is seen that the movement amount ratio (SY/sY) of the Y mass member 23 and the Y stage base 9, constituting the stage system of this embodiment, and the movement amount ratio (SP/sP) of the Y mass member 23 and the Y stage base 9 of the gravity gravity compensating mechanism, comprising pulley unit 43, are equal to each other. Because the movement amount ratio is the same, the Y mass member 23 and the Y stage base 9, constituting the inertial force applying mechanism can be connected to the pulleys of the gravity compensating mechanism through the belt 41. As a result, with a simple structure, both the inertial force applying mechanism and the gravity compensating mechanism are accomplished in parallel. While in this embodiment the gravity compensating mechanism uses a belt, this is not a requisition. A wire may be used, for example, as long as a similar effect is attainable. When plural Y mass members 23 are used, preferably plural pulley units 43 for supporting the Y stage base 9 or Y mass member 23 may be used similarly. In this embodiment, in addition to the effect of an inertial force applying mechanism of the first embodiment, there is an effect of the gravity compensating means comprising belts and pulleys that the drive reaction force resulting from drive of the Y stage base in the Y direction is reduced by movement of the Y mass member and, consequently, application of vibration to the base and the stage base is reduced. Furthermore, since the Y stage base and the Y mass member move in opposite directions with respect to the Y direction in a proportion corresponding to an inverse of the mass ratio of the Y stage base and the Y mass member, there occurs substantially no shift of gravity center position of the whole stage system. Therefore, deformation of the base for supporting the stage base can be suppressed. Additionally, by compensation of weight of the Y stage through pulleys and belts, the stage drive can be made with less energy and less heat generation. [Embodiment 3] FIG. 5 is a schematic and front view of a vertical type stage system according to a third embodiment of the present invention. The components corresponding to those of the first embodiment of FIG. 1 are denoted by the same reference numerals. Like the first embodiment, the third embodiment is applied to a vertical type X-Y stage system which is equipped with an inertial force applying mechanism and a gravity compensating mechanism. The principle of operation and the manner of operation of the gravity compensation mechanism of the second embodiment, comprising a cylinder mechanism, are essentially the same as those of the first embodiment, and a description therefor will be omitted. In the third embodiment, in place of driving a main stage and an X mass member independently through linear motors mounted on them, respectively, the main stage and the X mass member are provided with an X magnet unit and an X coil unit by which driving in the X direction and inertial force application are accomplished. Also, as regards the Y direction, a Y magnet unit and a Y coil unit are provided on the Y stage base and the Y mass member, by which driving in the Y direction and inertial force application are accomplished. Denoted in FIG. 5 at 1 is a base for supporting the vertical type stage system, and denoted at 2 is an anti-vibration damper for removing or reducing vibration of the base 1. Denoted at 3 is a stage base supported by the base 1, and by this stage base the stage unit is fixed. Denoted at 4 is a wafer chuck for holding a wafer thereon, and denoted at 5 is a main stage (first stage) for supporting the wafer chuck 4 and being movable along the wafer surface (X-Y plane). The main stage 5 has an X measurement mirror 6 and Y measurement mirror 7 fixedly mounted at an end face thereof, for providing reflection surfaces for a position measurement beam. Denoted at 8 is a Y stage guide fixedly mounted on the stage base 3, and denoted at 9 is a Y stage base (second stage) being movable in the Y direction (second direction) with the main stage 5 mounted thereon. The Y stage guide 8 serves to support the Y stage base 9 with respect to the X direction (first direction) and to guide it with respect to the Y direction without contact thereto. Denoted at 10 is an X stage guide fixedly mounted on the Y stage base 9. It serves to support the main stage 5 with respect to the Y direction and to guide it with respect to the X direction without contact thereto. This embodiment uses an air bearing mechanism for non-contact guiding. However, any other mechanism may be used, as long as it can provide a function of low friction guiding. Denoted at 21 is a Y magnet unit (magnet of a first driving mechanism) mounted on the Y stage base 9. Denoted at 22 is a Y coil unit (second coil) mounted on the Y mass member 23 (countermass). Here, the Y magnet unit 21 and the Y coil unit 22 provide a linear motor (second mechanism) which functions as moving means. Thus, the Y magnet unit and the Y coil unit may be disposed reversely. The Y mass member 23 is a mass member which is movable in an opposite direction to the Y stage base 9 with respect to the Y direction. Denoted at 24 is a Y mass guide fixedly mounted on the stage base, for guiding the Y mass member 23 in the X direction without contact thereto. As the Y mass member 23 moves in the Y direction, it functions as a Y-direction inertial force applying mechanism. Denoted at 25 is an X magnet unit (magnet of a second driving mechanism) mounted on the main stage 5. Denoted at 26 is an X coil unit (coil of second driving means) mounted on the X mass member 27 (secondary countermass). Here, the X magnet unit 25 and the X coil unit 26 provide a linear motor (second mechanism) which functions as moving means. Thus, the X magnet unit and the X coil unit may be disposed reversely. The X mass member 27 is a mass member which is movable in an opposite direction to the main stage 5 with respect to the X direction. Denoted at 28 is an X mass guide fixedly mounted on the Y stage base 9, for guiding the X mass member 27 in the X direction without contact thereto. As the X mass member 27 moves in the X direction, it functions as an X-direction inertial force applying mechanism. Now, driving the main stage 5 in the X direction with the structure described above, will be explained. As a driving signal is applied to the X coil unit 26 through a drive controller (not shown), X direction thrust is produced between the X coil unit 26 and the X magnet unit 25. Since, however, the main stage on which the X magnet unit 25 is mounted and the X mass member 27 on which the X coil unit is mounted are both arranged to be movable in the X direction and, additionally, since they are guided without contact to minimize the friction, they move in opposite directions with respect to the X direction. Here, the resultant force of thrust produced from the X magnet unit 25 for driving the main stage 5 is denoted by FX, while the a resultant force of thrust produced from the X coil unit 26 for driving the X mass member 27 is denoted by RX. From the relation of balance of force, the following equation applies. FX=xe2x88x92RX Here, if the X magnet unit 25 or the X coil unit 26 is arranged so that the operational axis of RX extends through the gravity center position of the main stage 5, then FX (main stage driving thrust) operates to pass through the gravity center position of the main stage 5. Thus, no rotational torque is produced at the main stage 5. Further, providing plural X mass members will be effective to broaden the latitude for designing the structure so that the operational axis of the drive thrust applied to the main stage 5 extends through the gravity center position. However, use of plural X mass members is not a requisition. Here, when the mass of the main stage 5 is MX and the total mass of the X mass member 27 is mX, the driving stroke sX for the X mass member 27 can be determined, with respect to the X-direction stroke SX of the main stage, by the mass ratio between MX and mX. This can be expressed by an equation below. SX/sX=1/(MX/mX) Namely, the ratio of movement amount of them displaces relative to the stage base 3, in a proportion corresponding to an inverse of the mass ratio. Thus, by making mX larger, the mass ratio MX/mX becomes smaller and, therefore, the driving stroke sX of the X mass member can be designed small. Since, however, the Y-direction movement mass including the X mass member 27 becomes larger, the energy required for Y-direction movement can be made smaller. The same applies to the Y direction. That is, as the Y stage base is driven in the Y direction, the Y mass member is driven. However, there is a large influence of gravity, in regard to the Y direction. Thus, here, a description will be made only on the function of Y-direction driving means, similarly to the case of the X direction, on an assumption that a gravity compensating mechanism acting on the Y stage base 9 or Y mass member 23 is being operated such that the effect of gravity applied to the Y stage base 9 or the Y mass member 23 can be disregarded. As a driving signal is applied to the Y coil unit 22 through a drive controller (not shown), Y direction thrust is produced between the Y coil unit 22 and the Y magnet unit 21. Since, however, the Y stage 9 on which the Y magnet unit 21 is mounted and the Y mass member 23 on which the Y coil unit 22 is mounted are both arranged to be movable in the Y direction and, additionally, since they are guided without contact to minimize the friction, they move in opposite directions with respect to the Y direction. Here, the resultant force of thrust produced from the Y magnet unit 21 for driving the Y stage base 9 is denoted by FY, while the resultant force of thrust produced from the Y coil unit 22 for driving the Y mass member 23 is denoted by RY. From the relation of balance of force, the following equation applies. FY=xe2x88x92RY Here, if the Y magnet unit 21 or the Y coil unit 22 is arranged so that the operational axis of RY extends through the gravity center position of the whole Y direction movement components, including the main stage 5 and Y stage base 9, then FY operates to pass through the gravity center position of the whole Y direction movement components. Thus, no rotational torque is produced at the stage. Further, providing plural Y mass members 23 will be effective to broaden the latitude for designing the structure so that the operational axis of the drive thrust applied to the Y stage base 9 extends through the gravity center position. However, use of plural Y mass members 23 is not a requisition. Here, when the mass of the Y stage base 9 is MY and the total mass of the Y mass member 23 is mY, the driving stroke sY for the mass member can be determined, with respect to the Y-direction stroke SY of the Y direction movement elements, by the mass ratio between MY and mY. This can be expressed by an equation below. SY/sY=1/(MY/mY) Namely, the ratio of movement amount of them displaces relative to the stage base 3, in a proportion corresponding to an inverse of the mass ratio. Thus, by making mY larger, the mass ratio MY/mY becomes smaller and, therefore, the driving stroke sY of the Y mass member can be designed small. In this embodiment, with the stage movement, the mass member is moved in an opposite direction to the stage. However, the relation between the movement amount of the stage and the movement amount of the mass member is similar to that of the first embodiment. Therefore, in the third embodiment, as in the first embodiment, the Y stage base 9 and the Y mass member 23, constituting the inertial force applying mechanism, can be connected to the gravity compensation mechanism through the cylinder rods A31 and B33. Thus, the inertial force applying mechanism and the gravity compensation mechanism can be accomplished in parallel, with a simple structure. Here, the ratio of sectional area between the pistons A32 and B34 is substantially the same as the mass ratio between the Y stage base 9 and the Y mass member 23. In this embodiment, a drive reaction force due to drive of the main stage in the X direction is reduced by movement of the X mass member and, as a result, application of vibration to be applied to the base or the stage base can be reduced. Additionally, since the main stage and the X mass member move in opposite directions with respect to the X direction, at a proportion corresponding to an inverse of the mass ratio, advantageously there occurs substantially no shift of gravity center position of the whole stage system. Therefore, deformation of the base which supports the stage base can be suppressed. Particularly, in a vertical type stage such as of the present embodiment, suppression of a shift of gravity center position or of a change in reaction force with respect to the X direction, is very effective to a high speed and high precision stage system, since it reduces a change in load applied to an anti-vibration damper. Further, in this embodiment, a reaction force resulting from drive of the Y stage base in the Y direction is reduced by movement of the Y mass member and, as a result, application of vibration to be applied to the base or the stage base can be reduced. Additionally, since the Y stage base and the Y mass member move in opposite directions with respect to the Y direction, at a proportion corresponding to an inverse of the mass ratio, advantageously there occurs substantially no shift of gravity center position of the whole stage system. Therefore, deformation of the base which supports the stage base can be suppressed. Furthermore, since the weight of the Y stage base is compensated for by a cylinder mechanism of a coupled structure, both the inertial force applying mechanism and the gravity compensating mechanism are accomplished in parallel, with a simple structure. Thus, stage driving can be performed with less energy and less heat generation. Further, in a case of an X-Y stage having a stage base reference surface extending horizontally, unlike the vertical stage of the present embodiment, the stage or mass member driving mechanism of this embodiment can be provided. On that occasion, the gravity compensating means of this embodiment may be omitted. [Embodiment 4] FIG. 6 is a schematic and front view of a vertical type stage system according to a fourth embodiment of the present invention. The components corresponding to those of the second embodiment of FIG. 3 or of the third embodiment of FIG. 5 are denoted by the same reference numerals. Like the preceding embodiment, the fourth embodiment is applied to a vertical type X-Y stage system which is equipped with an inertial force applying mechanism and a gravity compensating mechanism. The principle of operation and the manner of operation of the inertial force applying mechanism of the fourth embodiment, comprising a magnet unit and a coil unit, are essentially the same as those of the third embodiment, and a description therefor will be omitted. Further, the principle of operation and the manner of operation of the gravity compensation mechanism, comprising a pulley unit 43 and a belt 41, are essentially the same as those of the second embodiment, and a description therefor will be omitted. In this embodiment, a reaction force due to drive of the main stage in the X direction is reduced by movement of the X mass member and, as a result, application of vibration to be applied to the base or the stage base can be reduced. Additionally, since the main stage and the X mass member move in opposite directions with respect to the X direction, at a proportion corresponding to an inverse of the mass ratio, advantageously there occurs substantially no shift of gravity center position of the whole stage system. Therefore, deformation of the base which supports the stage base can be suppressed. Particularly, in a vertical type stage such as that of the present embodiment, suppression of a shift of gravity center position or of a change in reaction force with respect to the X direction, is very effective to a high speed and high precision stage system, since it reduces a change in load applied to an anti-vibration damper. Further, in this embodiment, a reaction force resulting from drive of the Y stage base in the Y direction is reduced by movement of the Y mass member and, as a result, application of vibration to be applied to the base or the stage base can be reduced. Additionally, since the Y stage base and the Y mass member move in opposite directions with respect to the Y direction, at a proportion corresponding to an inverse of the mass ratio, advantageously there occurs substantially no shift of gravity center position of the whole stage system. Therefore, deformation of the base which supports the stage base can be suppressed. Furthermore, since the weight of the Y stage base is compensated for by a gravity compensation mechanism comprising belts and pulley units, both the inertial force applying mechanism and the gravity compensating mechanism are accomplished in parallel, with a simple structure. Thus, stage driving can be performed with less energy and less heat generation. [Embodiment 5] FIG. 7 is a schematic and front view of a vertical type stage system according to a fifth embodiment of the present invention. The components corresponding to those of the first embodiment of FIG. 1 are denoted by the same reference numerals. Like the first embodiment, the fifth embodiment is applied to a vertical type X-Y stage system which is equipped with an inertial force applying mechanism and a gravity compensating mechanism. The principle of operation and the manner of operation of the inertial force applying mechanism of the fifth embodiment are essentially the same as those of the first embodiment, and a description therefor will be omitted. This embodiment uses a cylinder mechanism with a piston, for the gravity compensation mechanism. However, as compared with the embodiment described hereinbefore, a piston for supporting the Y stage base and a piston for supporting the Y mass member are provided by independent cylinder mechanisms, such that gravity compensation for these members is performed separately. Denoted in FIG. 7 at 31 is a cylinder rod (rod A) connected to the Y stage base. An end thereof opposite to the Y stage base 9 is supported by a cylinder piston A32. Denoted at 33 is a cylinder rod (rod B) connected to the Y mass member 23. An end thereof opposite to the Y mass member 23 is supported by a cylinder piston B34. Piston A of the gravity compensation mechanism (stage gravity compensation mechanism) for performing gravity compensation for the Y stage base 9, provides a sealing function to the fluid within an air cylinder A36. A gravity compensation controller (not shown) calculates a thrust level to be applied to the cylinder A36, and a thrust level signal is applied to a control valve (not shown) of the cylinder. On the basis of an electric current corresponding to the designated thrust level, the control valve operates to change the pressure or volume of the fluid within the cylinder A, to thereby adjust the thrust of the piston A32. The piston A32 is connected to the cylinder rod A31, and it applies a Y direction thrust to the stage base 9. Through the control of thrust of the cylinder mechanism 36, the weight of the Y stage base 9 is balanced with the thrust of the cylinder A36. Similarly, the piston B34 of the gravity compensation mechanism for performing gravity compensation for the Y mass member 23, provides a sealing function to the fluid within an air cylinder B37. On the basis of a thrust level designated by a gravity compensation controller (not shown), a control valve is controlled to change the pressure or volume of the fluid within the cylinder B, to thereby adjust the thrust of the piston B37. In this case, the weight of the Y stage base 9 is balanced with the thrust of the cylinder B37. As a result, gravity compensation for the Y stage base 9 and the Y mass member is accomplished. The cylinder A and piston A as well as the cylinder B and piston B are maintained without contact, to assure movement with very low friction. In this embodiment, the gravity compensation for the Y stage base and the gravity compensation for the Y mass member can be performed independently of each other. Therefore, the thrust of the piston of the cylinder mechanism performing the gravity compensation may be used as an energy for drive. Since the Y stage base and the Y mass member can be driven independently of each other, in addition to reduction of stage reaction force by the inertial force applying mechanism, there is an advantage that any vibration produced at the Y stage base due to external disturbance can be canceled. Further, there is no limitation to the sectional area ratio between the pistons A and B, as in the first embodiment. The latitude of design can be expanded more. Furthermore, it is possible to meet a change in mass of the Y stage base, when a wafer is placed on the stage. [Embodiment 6] FIG. 8 is a schematic and front view of a vertical type stage system according to a sixth embodiment of the present invention. The components corresponding to those of the third embodiment of FIG. 5 or of the fifth embodiment of FIG. 7 are denoted by the same reference numerals. Like the preceding embodiment, the sixth embodiment is applied to a vertical type X-Y stage system which is equipped with an inertial force applying mechanism and a gravity compensating mechanism. This embodiment comprises an X direction (second direction) inertial force applying mechanism having an X magnet unit 25 (magnet of a second driving mechanism) and an X coil unit 26 (coil of a second driving mechanism) mounted on the main stage 5 (second stage) and the X mass member 27 (secondary countermass), respectively, as well as a Y direction (first direction) inertial force applying mechanism having a Y magnet unit 21 (magnet of a first driving mechanism) and Y coil unit 22 (coil of a first driving mechanism) mounted on the Y stage base 9 (first stage) and the Y mass member 23 (countermass), respectively. The principle of operation and the manner of operation of these inertial force applying mechanisms are essentially the same as those of the third embodiment, and a description therefor will be omitted. In accordance with this embodiment, in addition to the advantageous effects of the third embodiment, there is an advantage that, in regard to the pistons A and B for supporting the Y stage base and the Y mass member, there is no limitation to the sectional area ratio therebetween. Thus, the latitude of design can be expanded more. Furthermore, it is possible to meet a change in mass of the Y stage base, when a wafer is placed on the stage. [Embodiment 7] FIG. 9 illustrates a stage system according to a seventh embodiment of the present invention. The stage system comprises an X-Y stage which includes: a Y stage 120 being reciprocally movable in the Y-axis direction (vertical direction or a direction close to it), along a stage base 110 mounted on a base plate (not shown); an X stage 130 being reciprocally movable in the X-axis direction along the Y stage 120; a pair of Y linear motors 140 which serve as first driving means for moving the Y stage 120 in the Y-axis direction; and an X linear motor 150 which serves as second driving means for moving the X stage 130 in the X-axis direction. In the illustration of FIG. 9, for explanation of a Y guide 111 (to be described later), one of the Y linear motors 140 placed at the left-hand side is omitted. The stage base 110 has an X-Y guiding surface 110a which provides a reference surface, extending in a vertical direction or a direction close to it, for non-contact support of the bottom face of the X stage 130 and the Y stage 120 through a static bearing device (not shown) having air pads, for example. At an end of the stage base 110 in the X-axis direction, there is a Y guide 111 (broken line) which serves as a yaw guide for guiding the Y stage 120 in the Y-axis direction. The Y guiding surface 111a of the Y guide 111 and the Y stage 120 are kept out of contact with each other, by means of air pads 120a, for example, of the yaw guide static bearing device. When both of the Y linear motors 140 are actuated, the Y stage 120 moves along the Y guide 111, over the X-Y guiding surface 110a of the stage base 110. The Y stage 120 is provided by a frame-like structure comprising a pair of Y sliders 121 and 122 as well as an X linear motor stator 152 having opposite ends supported by the sliders. The bottom faces of these Y sliders 121 and 122 are opposed to the X-Y guiding surface 110a of the stage base 110, and they are supported without contact, by means of the air pads, for example, as described above. The length of the Y slider 122 on the left-hand side as viewed in the drawing, is longer than the other. Its side face 122a is opposed to the Y guiding face 111a of the Y guide 111, and it is guided without contact, by use of air pads 120a (see FIG. 10B), for example, as described above. The Y sliders 121 and 122 are integrally connected to a Y linear motor movable element 141, through a connecting plate 123. The X stage 130 comprises a hollow plate member having a top plate 131. An X linear motor stator 152 extends through this hollow portion of the X stage. The top face of the top plate 131 provides a work stage for holding, by attraction, a wafer (workpiece) not shown. The Y linear motors 140 include Y linear motor movable elements 141, coupled integrally to the Y sliders 121 and 122, respectively, of the Y stage 120 through the connecting plate 123 as described, as well as Y linear motor stators 142 extending through apertures of them. In response to an electric current applied to the Y linear motor stators 142, a thrust in the Y-axis direction is produced at respective Y linear motor movable elements 141, by which the Y stage 120 and the X stage 130 are moved in the Y-axis direction. The X linear motor movable element for moving the X stage 130 in the X-axis direction, along the Y stage 120, is fixedly mounted on the inside face of the top plate 131. In response to an electric current applied to the X linear motor stator 152, a thrust in the X-axis direction is produced at the X linear motor movable element, by which the X stage 130 is moved in the X-axis direction and along the Y linear motor stator 152. A countermass mechanism 160 serves as a weight compensation mechanism for cancelling the weight of the Y stage 120, X stage 130 and the like. It comprises belts 162 which are connecting members having the Y sliders 121 and 122 (i.e., Y stage 120) suspended at one end and having a countermass 161 suspended at the other end, as well as pulleys 163 around which the belts are wound and supported. The weight of the countermass 161 is set so that it is balanced with the total weight of the whole stage movable components, including the Y stage 120, the X stage 130 and a wafer held thereby. As the X stage 130 moves in the X-axis direction, the gravity center position of the stage movable elements including the Y stage 120 and the X stage 130 displaces. Thus, the balance of rotational moment about the Z axis (wZ-axis direction) changes. However, this moment cannot be supported only by t he countermass mechanism 160, and an excessively large load is applied to the Y guide (yaw guide) 111 of the Y stage 120. In order to support such a large load, the rigidity of the Y guide 111 has to be enlarged. However, increasing the rigidity of the Y guide 111 necessitates enlargement in size of the Y guide 111, for example. This results in further enlargement in size and weight of the stage mechanism as a whole, and causes degradation of the dynamic characteristic of the stage. Finally, improvement of positioning precision or positioning speed is disturbed. For the belt that connects the Y stage 120 and the countermass 161, generally a steel belt or steel wire is used. When the Y stage 120 moves, there occurs natural or proper vibration due to insufficiency of rigidity of the steel belt or the like. In addition to this, natural vibration of the countermass 161 itself is propagated through the belt 62 to the Y stage 120. These vibrations degrade the stage positioning precision considerably, and this is a large bar to improvement of the frequency response characteristic of the positioning control system. In order to avoid vibration propagation from the belts 162 to the Y stage 120 and also to totally solve the problems attributable to rotational moment as described, actuators 170 are provided at connections between the Y stage 120 and the belts 162, which actuators have a function as anti-vibration means and which are operable to adjust the tension or effective length of each belt 162 in accordance with the displacement of the X stage 130. As shown in FIGS. 13-15, each actuator 170 may comprise a bellowphragm (air spring) 171 for controlling the tension of the belt 162, an air cylinder 172, or a linear motor 173, for example. Alternatively, it may comprise a laminated or accumulated rubber member 174 (FIG. 16) which is an elastic member for resiliently coupling the belt 162 and the Y stage 120 with each other. The drive amounts of the actuators 170 for respective belts 162, from which the Y sliders 121 suspend, are controlled independently of each other on the basis of the positional information regarding the X stage 130. By this, tension or effective length of each belt 162 can be adjusted. In this manner, a rotational moment produced with the movement of the X stage 130 can be canceled or compensated for and, by this, loads to be applied from the Y stage 120 to the Y guide 111 can be reduced. The bellowphragm 171, the air cylinder 172 or the linear motor 173 has an anti-vibration function, and thus it serves to absorb and attenuate natural vibration resulting from insufficient rigidity of the belt 162, or natural vibration of the countermass 161. Namely, by means of the actuators 170, vibration propagation from the belts 162 to the Y stage 120 can be prevented, such that the positioning precision or frequency response characteristic of the positioning control system can be improved significantly. The laminated rubber member 174 of FIG. 16 may not have an effect of rotational moment correction to the Y stage 120. However, it provides a function for effectively absorbing vibration to be propagated from the belt 162 to the Y stage 120. As described above, the guiding face 111a of the Y guide 111 and the Y stage 120 (Y slider 122) opposed to it are kept without contact with each other, by means of the air pads 120a. In addition to the air pads 120a, the Y stage 120 is equipped with magnetic pads 120b which serve to apply a preload in an opposite direction to the air pads 120a, to thereby produce bearing rigidity k1 shown in FIG. 10A. Further, as shown in FIG. 11, in parallel to the Y guide 111, there is a countermass yaw guide 164 at the back of the stage base 110. The countermass yaw guide 164 is disposed opposed to air pads 161a and magnetic pads 161b which serve as a countermass yaw guide static bearing device, provided at one end of the countermass 161. By means of these pads, the countermass 161 is guided in the Y-axis direction, without contact. The magnetic pads 161b of the countermass 161 serve to apply a preload in an opposite direction to the air pads 161a, whereby a bearing rigidity k2 (FIG. 10A) which is larger than the bearing rigidity k1 on the Y stage 120 side, is produced. The bearing rigidity k2 of the countermass yaw guide 164 is made larger than the bearing rigidity k1 of the Y guide 111 as described. This is to assure that the influence of rotational moment, produced as the gravity center position of the Y stage 120 shifts with movement of the X stage 130 in the X-axis direction, is received by the countermass 164 to thereby reduce the rotational moment to be applied to the Y stage 120. The position of the X stage 130 with respect to the Y-axis direction and X-axis direction can be measured by using position sensors 130c and 130d, respectively, for receiving reflection light from a Y measurement mirror 130a and an X measurement mirror 130b, integral with the X stage 130. Next, a case where each actuator 170 comprises a bellowphragm 171 shown in FIG. 13, will be explained. The bellowphragm 171 includes a bellows 171b having an air supply port 171a. The upper end of the bellows 171b is coupled to a first housing 171c which is integral with the Y stage 120. The lower end of the bellows 171b is connected to a second housing 171d, connected to the lower end of the belt 162. By changing the pressure of air to be supplied to the air supply port 171a, the air pressure within the bellows 171b is changed, such that the tension of the belt 162 is changed. Through the damper effect of the air pressure within the bellows 171b, vibration propagation from the belt 162 to the Y stage 120 can be effectively reduced or prevented. FIG. 17 is a block diagram of a control system for controlling the inside air pressure of the bellows 171b, on the basis of positional information of the X stage 130 in the X-axis direction. A servo system for controlling X linear motor 150 controls the drive amount of the X linear motor 150, on the basis of a position designation value supplied thereto from a computer (not shown) and of positional information of the X stage 130 in the X-axis direction as fed back from the position sensor 130d. The position designation value described above is kp (proportion) transformed, and the result is supplied to a controller 170a (control means) together with a pressure designation value of a bellowphragm control system. In response, a servo valve connected to the air supply inlet 171a of the bellowphragm 171 is adjusted, whereby the inside air pressure within the bellows 171b is controlled. The position designation value of the X stage 130 is multiplied by a conversion coefficient kp and a pressure designation value of the bellowphragm control system is added. Then, as the gravity position of the stage movable portion shifts, inside pressure of each bellows 171 is changed to adjust the tension of each belt 162. In this manner, the tensions of belts 162 are adjusted individually. This produces a rotational moment opposite to the rotational moment, produced with the movement of the X stage 130. As a result, loads to be applied to the Y guide 111 are reduced. Since the load applied to the Y guide 111 can be reduced remarkably through the moment correction described above, a dynamic characteristic of the stage can be improved largely without enlargement in size or weight of the Y guide 111. It is, therefore, possible to meet improvement of speed and precision of stepwise motion and positioning. Next, an anti-vibration effect of the actuator 170 comprising bellowphragm 171, air cylinder 172 or linear motor 173, for example, or of an elastic member such as a laminated rubber member 174, will be explained in greater detail. In a positioning control system for controlling Y linear motor 140 or X linear motor 150 for driving the Y stage 120 or X stage 130, generally a PID (Proportion Kp, Integration Ki and Differentiation Kd) operation is performed, and, as shown in the graph of FIG. 19, gain adjustment is made so that the gain crossover frequency with which the servo rigidity xcex94d/xcex94e becomes lowest is set to about 100 Hz. In a case where no actuator or no laminated rubber member is added, if, as a result of resonance of a belt and a countermass suspending therefrom, for example, vibration of 30 Hz enters the positioning control system as external disturbance, then, because of low servo rigidity about several tens of Hz (FIG. 19) the work stage for holding a wafer will be oscillated largely. On the other hand, if vibration at several tens of Hz propagated from the belt is removed by using anti-vibration means such as an actuator or laminated rubber, only vibration of 1 Hz, for example, will be transmitted to the work stage. If there is only an external disturbance of a low frequency as a few Hz, because of high servo rigidity of the positioning control system, the dynamic characteristic would not be degraded. Improving the dynamic characteristic of a small size and high performance stage system largely as described above, significantly contributes to reduction in size and enhancement of performance and productivity of an exposure apparatus for the manufacture of semiconductor devices, for example. When each actuator 120 comprises an air cylinder 172 (FIG. 14) in place of bellowphragm 171, the following applies. The cylinder 172b having an air supply port 172a is made integral with the Y stage 120. The belt 162 has its lower end connected to the piston 172c. By changing the pressure of air to be supplied to the air supply inlet 172a, the tension of the belt 162 is adjusted. A control system for controlling the air pressure of the cylinder 172b may be similar to that shown in FIG. 17. When each actuator 170 comprises a linear motor 173 shown in FIG. 15, the following applies. The coil 173a of the linear motor 173 is integral with the Y stage 120, and the belt 162 has its lower end connected to a driving magnet 173b. By changing an electric current to be applied to the coil 173a, the tension of the belt 162 is adjusted. A control system for controlling the electric current to be supplied to the coil 173a may be similar to that of FIG. 17. The laminated rubber member 174 shown in FIG. 16 has its upper end connected to a housing 174a which is integral with the Y stage 120. The lower end of the rubber member 174 is supported by a housing 174b, connected to the belt 162. Through elastic change in thickness of the laminated rubber member 174, vibration propagation from the belt 162 to the Y stage 120 is prevented. In the structure shown in FIG. 16, the laminated rubber member 174 may be replaced by a piezoelectric device which can function as an actuator. On that occasion, the housing 174a supporting the upper end of the piezoelectric device may be integral with the Y stage 120, while a housing 174b for supporting the lower end of the piezoelectric device 174 may be connected to the lower end of the belt 162. By changing an electric voltage to be applied to the piezoelectric device, the thickness thereof is changed to adjust the effective length of the belt 162. A control system for controlling the electric voltage to be supplied to the piezoelectric device may be similar to that of FIG. 17. It is to be noted that, even through rotational moment correction of the Y stage 120 by actuators 170 or through reduction of vibration propagation to the Y stage 120 through belts 162, it may be still difficult to completely avoid vibration propagation from the countermass 161 to the Y stage 120. While such vibration may be of a low frequency as a few Hz, for further improvement of positioning precision and for attainment of higher speed, it may be undesirable. In consideration of this, third driving means comprising a pair of linear motors 180 (FIG. 12) may preferably be added to accelerate the countermass 161 in an opposite direction to the Y stage 120, to thereby suppress the natural vibration of the countermass 161. The linear motors 180 may be provided at opposite ends of the countermass 161, and they may be disposed at the back of the Y linear motor 140 for moving the Y stage 120 in the Y-axis direction. Each linear motor 180 is provided with a linear motor movable element 181 integral with the countermass 161, which element is movable along a linear motor stator 182 provided at a side edge of the stage base 110. By controlling electric current to be supplied to the linear motor stator 182, acceleration of the countermass 161 in the Y-axis direction can be adjusted to one having the same absolute value as and in an opposite direction to the acceleration as provided by the Y linear motor 140 to the Y stage 120. By doing so, it is possible to substantially completely remove external disturbance due to natural vibration of the countermass 161. A control system for the linear motor 180 for driving the countermass 161 will be described later with reference to the block diagram of FIG. 18. Through attenuation of vibration propagation to the Y stage 120 from the countermass mechanism 160 including belts 162 by the actuators 170 and through reduction of natural vibration of the countermass 161, external disturbance of the control system of the Y stage 120 can be removed very effectively. This contributes to further improvement of positioning precision and positioning speed. FIG. 18 shows a control system for linear motors 180, for driving the countermass 161 in the Y-axis direction. A servo system for controlling the Y linear motor 140 controls the driving amount of the Y linear motor 140 on the basis of a position designation value as supplied from a computer (not shown) and of positional information of the Y stage 120 in the Y-axis direction as fed back from the position sensor 130c. Also, the position designation value is applied to the linear motor 180 of the countermass 161, as an acceleration designation value being twice differentiated and kp converted. Only with the addition of such a simple control system as described above, the driving amounts of the Y stage 120 and of the countermass 161 can be controlled synchronously. [Embodiment 8] FIG. 20 shows an eighth embodiment of the present invention. In this embodiment, in place of provision of actuators 170 at the connections between the Y stage 120 and the belts 162, actuators 180 are provided at the bearing portion (supporting portion) of pulleys 163 disposed at the top of the stage base 110. As shown in FIG. 21, each actuator 180 comprises bellowphragms 181 provided between the stage base 110 and a bearing base 163b on which a rolling bearing 163a for rotatably supporting the pulley 163 is mounted. The inside structure of the bellowphragm 181 may be similar to that of the bellowphragm 171 of FIG. 13. It is controlled with a control system similar to that of FIG. 17. In place of the bellowphragm 181, an air cylinder, a linear motor, or a laminated rubber member such as shown in FIGS. 14-16 may be used. As regards stage base 110, Y guide 111, Y stage 120, X stage 130, Y linear motor 140, X linear motor 150, countermass mechanism 160 and so on, they may be similar to those of the seventh embodiment. A description therefor will be omitted, by assigning corresponding reference numerals to them. As described above, an actuator or elastic member (laminated rubber member) for adjusting the tension or effective length of the belt may be provided at the support of the pulley. Also, similar actuator means may be provided at the connection between the countermass and the belt at the back of the stage base, with a similar advantage of effective correction or prevention of rotational moment. As a further alternative, in place of the rolling bearing 163a of FIG. 21, a rotary type static pressure bearing 190 shown in FIG. 22 may be used to rotatably support the pulley 163. The rotary type static bearing 190 comprises a pair of bearing members 190a for supporting the bearing portion of the pulley 163 with respect to a radial direction and a thrust direction, without contact thereto. It serves to resiliently change the height of the pulley 163 within the range as determined by the bearing gap 190b of the bearing member 190a. With the damper effect as described, vibration propagation from the belt 162 to the Y stage 120 can be prevented. [Embodiment 9] FIG. 23 shows a ninth embodiment of the present invention. In this embodiment, in place of the countermass 161 of the seventh and eighth embodiments, a weight compensation mechanism 200 is used wherein a pulley 263 is rotationally driven by a weight compensation motor 203 which produces a torque that balances with the weight of the stage movable portion, including the Y stage 220 and X stage 230. More specifically, each motor 203 rotates the pulley 263 to wind up the belt 262, to thereby cancel the weight of the stage movable portion. Each of the belts 262 and the Y stage 202 (Y slider 221) are connected to each other through an actuator 210 (anti-vibration means). As regards stage base 210, Y guide 211, Y stage 220, X stage 230, Y linear motor 240, X linear motor 250 and so on, they may be similar to those of the first embodiment, and a description therefor will be omitted. As for actuators 270, a bellowphragm 171, an air cylinder 172 or a linear motor 173 shown in FIGS. 13-15 may be used. In place of the actuator 210, an elastic member such as rubber member 174 of FIG. 16 or a piezoelectric device may be used. In a weight compensation mechanism without a countermass, if the Y stage is suspended by using a connecting member such as a belt, natural vibration of the belt will become a factor for deteriorating the dynamic characteristic of the stage. Thus, like the seventh or eighth embodiment, an actuator or elastic member having an anti-vibration function may be added. This largely contributes to improvement of speed and precision of the stage system. The actuators may be controlled on the basis of positional information about the X stage to perform rotational moment correction, similarly to the seventh and eighth embodiments. The actuator or elastic member (rubber member) may be provided at the connection between the Y stage and the belt, as shown in FIG. 23. Alternatively, it may be provided at the support for the bearing portion of the pulley. As a further alternative, an elastic member such as a laminated rubber member may be provided on the surface of a pulley for winding the belt, to thereby absorb belt vibration. [Embodiment 10] Next, an embodiment of an X-ray exposure apparatus which uses a stage system according to any one of the preceding embodiments, will be explained. FIG. 24 is a schematic view of an X-ray exposure apparatus according to an embodiment of the present invention. Synchrotron radiation light emitted from a synchrotron radiation producing device 301 (X-ray source) impinges on a mirror 302 disposed at a predetermined distance from the light emission point. The mirror 302 has a convex surface shape, and it expands the synchrotron radiation light of a sheet-like beam from the synchrotron radiation source 301. Although only one mirror is illustrated, plural mirrors may be used for expansion of the synchrotron radiation. The synchrotron radiation light reflected by the mirror passes through a transmission type mask M (original) having a pattern of an X-ray absorptive material, formed on an X-ray transmissive film, by which the light is transformed into a desired pattern shape and is projected on a substrate W (wafer). The wafer is coated with a resist material (photosensitive material). The wafer W is held by a wafer chuck 303 of a stage system, according to any one of the preceding embodiments of the present invention described above. The wafer chuck 303 is mounted on a main stage (not shown). Disposed upstream of the mask M is a shutter 304 for controlling exposure time over the whole exposure region. The shutter 304 is actuated by a shutter driving unit 304a which is controlled by a shutter control unit 304b. A beryllium film (not shown) is provided between the mirror 302 and the shutter 304. The mirror side of this beryllium film is maintained at a super vacuum, while the shutter side thereof is maintained at a reduced pressure He. In accordance with this embodiment of the present invention, an exposure apparatus which meets increases of speed and precision is provided. [Embodiment 11] Next, an embodiment of a device manufacturing method which uses an exposure apparatus such as described above, will be explained. FIG. 25 is a flow chart of a procedure for the manufacture of microdevices such as semiconductor chips (e.g., ICs or LSIs), liquid crystal panels, CCDs, thin film magnetic heads or micro-machines, for example. Step 1 is a design process for designing a circuit of a semiconductor device. Step 2 is a process for making a mask on the basis of the circuit pattern design. Step 3 is a process for preparing a wafer by using a material such as silicon. Step 4 is a wafer process which is called a pre-process wherein, by using the so prepared mask and wafer, circuits are practically formed on the wafer through lithography. Step 5 subsequent to this is an assembling step which is called a post-process wherein the wafer having been processed by step 4 is formed into semiconductor chips. This step includes an assembling (dicing and bonding) process and a packaging (chip sealing) process. Step 6 is an inspection step wherein an operation check, a, durability check and so on for the semiconductor devices provided by step 5, are carried out. With these processes, semiconductor devices are completed and they are shipped (step 7). FIG. 26 is a flow chart showing details of the wafer process. Step 11 is an oxidation process for oxidizing the surface of a wafer. Step 12 is a CVD process for forming an insulating film on the wafer surface. Step 13 is an electrode forming process for forming electrodes upon the wafer by vapor deposition. Step 14 is an ion implanting process for implanting ions to the wafer. Step 15 is a resist process for applying a resist (photosensitive material) to the wafer. Step 16 is an exposure process for printing, by exposure, the circuit pattern of the mask on the wafer through the exposure apparatus described above. Step 17 is a developing process for developing the exposed wafer. Step 18 is an etching process for removing portions other than the developed resist image. Step 19 is a resist separation process for separating the resist material remaining on the wafer after being subjected to the etching process. By repeating these processes, circuit patterns are superposedly formed on the wafer. With these processes, high density microdevices can be manufactured. While the invention has been described with reference to the structures disclosed herein, it is not confined to the details set forth and this application is intended to cover such modifications or changes as may come within the purposes of the improvements or the scope of the following claims. |
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claims | The ornamental design for an infrared sensing device, as shown and described. |
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051464810 | abstract | A substantially compressive stress-free, pin-holes free, and defects free continuous polycrystalline diamond membrane for an x-ray lithography mask is produced by placing a prepared substrate into a hot filament chemical vapor deposition reaction chamber, pre-heating the substrate to 400.degree. C.-650.degree. C. in the presence of an inert gas, heating the substrate to 650.degree. C.-700.degree. C. in the presence of hydrogen and carbon compounds, and chemically vapor depositing a polycrystalline diamond membrane onto the substrate. |
041464290 | summary | BACKGROUND OF THE INVENTION The field of this invention relates to nuclear reactors, and more particularly to an emergency disposal system for an "out of control" nuclear reaction, this disposal system taking the form of a mass separation system which separates the larger nuclear mass into a plurality of smaller masses. Nuclear reactors contain a plurality of fissionable material in the form of a plurality of elongated rods. Surrounding each of these rods is a control rod assembly and the nuclear reaction is controled by movement of the control rods and exposure of the fissionable material in one rod to the fissionable material in another rod. The rate of reaction is directly dependent upon the size of the mass. The more mass of fissionable material exposed, the greater the rate of reaction. If perchance an excessive amount of mass becomes exposed, what occurs is termed a "core meltdown". A "core meltdown" is perhaps the most discussed and most feared phenomenon of all potentialities of nuclear reactors. This problem is especially of major concern since the use of nuclear reactors to produce electricity is becoming quite common. The fantastic heat generated by a molten core, estimated about six thousand degrees Farenheit, is sufficient to melt through virtually any material that might be placed in its path as it bores downward through the reactor vessel and all containments, eventually releasing at least a portion of its radioactive fission product inventory through the earth, or possibly through breaks in the containment sphere. At the present time, the use of reliable and practical methods of containing a large molten mass of fuel that would probably result from a meltdown do not exist. At the present time, there has not been a complete meltdown of a nuclear reactor core. However, in certain instances there have been partial meltdowns which have been stopped by currently employed safety structure within the nuclear reactor. Even the occurrence of a partial meltdown has been described as a bit worse then the maximum credible accident. There is a chance that a complete meltdown would result in the fissionable material being directly exposed to the atmosphere and the pollution that would occur as a result thereof would undoubtedly be overwhelming. SUMMARY OF THE INVENTION The subject matter of this invention relates to structure which is to be incorporated within the building foundation for a nuclear reactor. If a complete meltdown occurs within a nuclear reactor, the molten mass of high temperature fissionable material will move by gravity in a downward direction, consuming everything in its path. The structure of this invention is to route the molten mass of material into a plurality of separate paths and then to further separate the separate paths of nuclear mass into still smaller separate quantities of fissionable material. By the separating of the combined nuclear mass into a plurality of separate masses, the rate of reaction is significantly slowed and also the produced heat of reaction is slowed. The net result is to eliminate the hazard produced by the combined quantity of nuclear mass. The structure of this invention achieves the prior objective by locating a plurality of passageways within the building foundation of a nuclear reactor facility. The passageways are constructed so that the molten nuclear mass is caused to be conducted within a plurality of separate first passageways and each first passageway is connected to a plurality of separate second passageways. Each second passageway is connected to a plurality of third passageways and, if needed, each third passageway is connected to a plurality of fourth passageways. At the junction between different passageways there may be included a readily meltable material, such as lead or tin or oxides thereof. The function of this material is to slow the movement of the nuclear mass and permit such to be evenly distributed between the different passageways. Also, the combining of the meltable material with the fissionable material functions to disperse the material and further decrease the rate of reaction. At each junction, the cross-sectional area of the passage leading to the junction is to be approximately equal to the cross-sectional area of the combined passageways leading from the junction. The fourth passageways may terminate in an explosive charge with this explosive charge being located in a bed of sand. Upon the fissionable material reaching the explosive charge, the explosive charge will automatically ignite causing the fissionable material to be finely dispersed within the sand. This fine distribution of the diffusable material would decrease the rate of reaction to where it is no longer considered harmful. |
052456455 | abstract | A structural part for a nuclear reactor fuel assembly includes a zirconium alloy material having at least one alloy ingredient selected from the group consisting of oxygen and silicon, a tin alloy ingredient, at least one alloy ingredient selected from the group consisting of iron, chromium and nickel, and a remainder of zirconium and unavoidable contaminants. The zirconium alloy material has a content of the oxygen in a range of substantially from 700 to 2000 ppm, a content of the silicon of substantially up to 150 ppm, a content of the iron in a range of substantially from 0.07 to 0.5% by weight, a content of the chromium in a range of substantially from 0.05 to 0.35% by weight, a content of the nickel of substantially up to 0.1% by weight, and a content of the tin in a range of substantially from 0.8 to 1.7% by weight. The alloy ingredients selected from the group consisting of iron, chromium and nickel are precipitated out of a matrix of the zirconium alloy as secondary phases, having a diameter with a geometric mean value in a range of substantially from 0.1 to 0.3 .mu.m. The degree of recrystallization of the zirconium alloy is less than or equal to 10% and a sample of the zirconium alloy, after a recrystallization annealing with a degree of recrystallization of 97.+-.2%, has a mean grain diameter less than or equal to 3 .mu.m. |
summary | ||
description | This patent document relates to systems, devices, and processes for thermonuclear fusion technologies. A tokamak is a device that uses a magnetic field to spatially confine plasma, e.g., in a shape of a torus to produce high-temperature plasma needed for producing controlled thermonuclear fusion power. Magnetic fields are used in tokamak devices for confinement in part because solid materials cannot withstand extremely high temperatures of the plasma for thermonuclear fusion. In a tokamak, stable plasma equilibrium can be achieved by producing magnetic field lines that move around the torus in a helical shape. Such a helical field can be generated by adding a toroidal field that travels around the torus in circles and a poloidal field that travels in circles orthogonal to the toroidal field. In implementations, the toroidal field can be produced by electromagnets that surround the torus, and the poloidal field can be produced by a toroidal electric current that flows inside the plasma and may be induced inside the plasma by, e.g., using a second set of electromagnets. At large toroidal currents, various magnetic confinement fusion devices tend to exhibit undesired plasma instabilities. The nonlinear evolution of such plasma instabilities can lead to a quench of the plasma current within a short period time, e.g., around milliseconds. This quench can create energetic runaway electrons that escape the spatially confined plasma and can potentially lead to a rapid loss of plasma confinement. Such runaway electrons can collide with components that face the plasma and can damage the components by, e.g., inflicting intense heat in such components. This phenomenon is called a plasma disruption. Techniques, systems, and devices are disclosed for exemplary pellets that can be used for magnetic fusion devices for mitigating plasma disruption. The exemplary pellets can mitigate plasma disruptions while penetrating the plasma. Thus, the disclosed technology can enhance the penetration depth of the pellets and can allow the pellets to reach the central core of the plasma. The exemplary pellets can also be introduced into the magnetic fusion devices in the form of a hollow shell comprising an interior payload of small granules or porous material, thereby diminishing the potential for damage to the far wall of the magnetic fusion device in the event of accidental impact. An exemplary embodiment discloses a fusion device. The exemplary fusion device comprises a plasma vessel structured to include a hollow interior to confine plasma, a plurality of toroidal field coils wound around different portions of an outer surface of the plasma vessel, the plurality of toroidal field coils configured to magnetically confine plasma inside the plasma vessel, a storage device that stores pellets, and a pellet injector positioned to receive pellets from the storage device and operable to inject the pellets into the plasma vessel. In some embodiments, the storage device is a cryostat storage device that stores and cools pellets, and wherein the pellets have metallic exteriors. In some embodiments, wherein the cryostat storage device is configured to cool the plurality of pellets to less than or equal to 40 kelvin (K). In some embodiments, the cryostat storage device is configured to cool the plurality of pellets to 10 kelvin (K). In some embodiments, the pellets include solid pellets. In some embodiments, the pellets include hollow shell pellets. In some embodiments, each hollow shell pellet encapsulates a payload. In some embodiments, the payload comprises granules or a porous material. In some embodiments, the payload comprises lithium, lithium deuteride, beryllium, beryllium deuteride, boron, boron nitride, or tungsten. In some embodiments, each pellet includes lithium or beryllium. In some embodiments, the plasma vessel is shaped as a D-shaped torus. In some embodiments, the pellet injector includes a single-stage light gas gun. In some other embodiments, each pellet includes a hollow shell that encapsulates a payload. In some other embodiments, the hollow shell comprises lithium, lithium deuteride, beryllium, beryllium deuteride, or boron nitride. In some other embodiments, the payload comprises lithium, lithium deuteride, beryllium, beryllium deuteride, boron, boron nitride, or tungsten. Another embodiment discloses a method of mitigating plasma disruption. The exemplary method comprises magnetically confining plasma in a plasma vessel, storing pellets, and injecting the stored pellets into the plasma vessel. In some embodiments, the exemplary method further comprises cooling the stored pellets to less than or equal to 40 kelvin (K). In some embodiments, the stored pellets are cooled to approximately 10 kelvin (K). Based on the disclosed technology in this document, in operating a magnetic confinement fusion device such as tokamak, pellets can be injected into hot fusion plasma to mitigate the undesired plasma disruption. However, the injected pellets tend to rapidly vaporize due to the heating caused by the hot fusion plasma. This vaporization can limit the depth of penetration of the injected pellets into the hot fusion plasma. Penetration of the pellets to the core of the confined plasma before completely vaporizing can mitigate an undesired plasma disruption. The interaction of the plasma with the ablated material of the pellets can cause the plasma to radiate away its thermal energy thereby spreading it over a wide area instead of allowing an uncontrolled loss of thermal energy to the surrounding plasma facing components of the fusion device. Plasmas can be confined by nested magnetic flux surfaces, but when these surfaces are broken during a plasma disruption, the thermal energy of the plasma can rapidly escape to the surrounding material structures causing damage. The disclosed technology in this document can be used to mitigate plasma disruptions based in part on engineering and control of the chemical composition, the structure, or the temperature of the exemplary pellets. In some embodiments, the structure of the exemplary pellets can include a shell and materials inside the shell known as payload. The shell and the payload once dispersed into the plasma can be ionized by the plasma and can allow energy to radiate out of the plasma. The exemplary pellets can aide in mitigating plasma disruption in part because the shell of the pellets can be used to carry at least some of the payload into the plasma before some portion of the payload begins to ionize. The exemplary embodiments disclose both cooled and non-cooled pellets that can differ in how far the shell can penetrate into the plasma before the shell disintegrates releasing both the shell material and the payload into the plasma. The exemplary embodiments first describe pellets that can be cryogenically cooled prior to introducing the pellet in a magnetic confinement fusion device. The cooled pellets may have solid forms or hollow shell forms. Each cooled pellet may include lithium or beryllium. The cooled pellet with a hollow shell form can encapsulate a payload. The payload of a cooled pellet may comprise granules or a porous material. The payload of a cooled pellet may comprise lithium, lithium deuteride, beryllium, beryllium deuteride, boron, boron nitride, or tungsten. The exemplary embodiments also describe pellets that may not be cryogenically cooled prior to introducing the pellet in a magnetic confinement fusion device. The pellet that is not cooled may include a hollow shell that encapsulates a payload. The hollow shell of the pellets that are not cooled may comprise lithium, lithium deuteride, beryllium, beryllium deuteride, or boron nitride. The payload of the pellets that are not cooled may comprise lithium, lithium deuteride, beryllium, beryllium deuteride, boron, boron nitride, or tungsten Turning to embodiments that use cryogenic treatment, examples of suitable pellets may include pellets having metallic exteriors which may include a metal with a low atomic number (Z). The exemplary pellets may optionally be placed inside a cryostat at a cryogenically low temperature that causes a rise in the electrical conductivity of the cooled pellets. The cooled pellets can be transferred to a pellet gun and accelerated to a desired high velocity into a magnetic fusion device, such as a tokamak. The high conductivity of the pellets can screen out the plasma's magnetic field from the interior of each pellet. This screening slows the ablation rate of the pellet allowing for deeper pellet penetration and a better suited spatial profile of deposited material for proper mitigation of the plasma disruption. FIG. 1 shows a cross section of an exemplary magnetic fusion device (100) where a cryogenically cooled pellets (116) are injected into a plasma vessel (115) that is used to spatially confine magnetized plasma (111) in the shape of a torus. As indicated by the center line (CL) on the left of FIG. 1, the plasma vessel (115) may be constructed as a D-shaped torus. The magnetic field region of the plasma core (111) is enclosed by a magnetic cage that includes multiple toroidal field coils (110). The magnetic cage may be formed by a plurality of toroidal field coils (110) that are wound around different portions of the outer surface of the plasma vessel. The magnetic cage is structured to magnetically confine the plasma (111). The cross section of the plasma (111), which forms a torus with axis of symmetry (CL) indicated by the dotted line on the left. The field may be initially excluded from the interior of the exemplary pellet (116) once the pellet crosses the magnetic cage. The white region within the final closed flux surface represents the plasma core (111). In some embodiments, a pellet injector (106) is used to inject the exemplary pellets (116) into the plasma (111). For example, the pellet injector (106) may include a single-stage light gas gun. The interaction of the plasma with the ablated material of an exemplary pellet (116) causes the plasma (111) to radiate away its thermal energy thereby spreading it over a wide area instead of allowing an uncontrolled loss of thermal energy to the surrounding plasma facing components. Plasmas (111) can be confined by nested magnetic flux surfaces (112). The same magnetic field that confines the plasma can also be manipulated to control the ablation rate of the pellet (116). FIG. 2 shows an exemplary embodiment where a magnetic screening (206) of an exemplary pellet (208) diverts the plasma's magnetic field (202) around the pellet. The plasma electrons (204) follow the magnetic field lines (202), which are excluded, from the interior of the metallic pellet (208) by the diamagnetic effect. Further, the heat flux impinging on the pellet (208) is carried by plasma electrons (204), which are strongly magnetized because they are pinned to the magnetic lines of force (202). Consequently, when these lines of force (202) are diverted around the pellet (208) the heat flux will also be diverted around the pellet. As a result the vaporization rate of the pellet can be reduced, thus allowing for deeper penetration of the pellet into the interior of the plasma. The physical mechanism that causes the magnetic field lines (202) to be diverted around a metallic pellet (208) is a diamagnetic phenomenon. The exemplary pellet (208) may be either in a solid form or in the form of a hollow shell of a suitable geometry such as a hollow spherical shell. The hollow interior of a hollow shell can be filled with a payload material. In some embodiments, the payload may be in the form of loosely packed granules, e.g., in the size range of 10 to 200 microns in some implementations. In some other embodiments, the payload may be in the form of a porous material. In some embodiments, the exemplary pellet (208) used in disruption mitigation may include a solid that includes a low-Z metal or light metal such as lithium or beryllium. In some embodiments, the exemplary pellet (208) used in disruption mitigation may include a hollow shell form where the outer shell material for the pellet is comprised of either lithium or beryllium. In an exemplary embodiment, the payload inside the hollow shell may include low-Z material, such as one or more of lithium (Li), lithium deuteride (LiD), beryllium (Be), beryllium deuteride (BeD), boron (B), boron nitride (BN), and tungsten (W). In some embodiments, the electrical conductivity of low-Z or light metal elements such as lithium or beryllium can be increased by 100 to 1000 times the room temperature value by cooling these materials down to approximately 10 kelvin (K) using a suitable cryostat storage device (102). In some embodiments, high electrical conductivity can be achieved for low-Z light metal pellets of lithium or beryllium for disruption mitigation in magnetic fusion device plasmas such in tokamaks by cooling pellets to 40 K or less using the cryostat storage device (102). Lithium and beryllium materials are non-magnetic crystalline solid metals for which the temperature dependence of the electrical conductivity arises from electron-phonon scattering. In this case the resistivity ρ (inverse of electrical conductivity, ρ=1/σ) is given by the following Block-Gruneisen formula: ρ = ρ ( 0 ) + αρ * ( T Θ D ) 5 ∫ 0 Θ D / T x 5 dx ( e x - 1 ) ( 1 - e - x ) where ρ (0) is the residual resistivity at zero temperature due to electron scattering from crystal defects or impurities, and the second term is the temperature dependent part arising from electron-phonon interaction, ΘD is the Debye temperature, ρ* is the resistivity at this temperature, and α=4.225 is a numerical constant. At temperatures well below the Debye temperature, typically approximately 300-400 K, the resistivity falls dramatically as T5 approaches zero. FIG. 3 shows a plot of electrical resistivity versus temperature for lithium sample, (Li) (302), and beryllium sample, (Be) (polycrystalline (304); pure crystal (306)). FIG. 3 shows that the electrical conductivity improves dramatically on cooling to low temperatures, especially near absolute zero, although the residual resistivity at zero temperature is apparent. The residual resistivity varies from sample to sample depending on how the sample is prepared and how pure it is. The high conductivity brought about by pre-cooling the pellets can make the diffusion time of the magnetic field into the pelletτd˜μ0σrp2 comparable to the penetration of the pellet with velocity Vτp˜a/V where σ is the electrical conductivity, μ0 is the permeability of free space equal to 4π×10−7, rp is the pellet radius, and a is the minor radius of the plasma. Returning to FIG. 1, the pellet (116) may be injected at some distance from the plasma (111) where there is no magnetic field. The pellet injector (106) is coupled to the plasma vessel (115). In some embodiments, the pellet injector (106) may inject the pellet (116) into the pellet injection tube (108). One end of the pellet injection tube (108) is connected to the pellet injector (106), and another end of the pellet injection tube (108) may be coupled to the plasma vessel (115) and may extend past the toroidal field coils (110) and the stainless steel (SST) walls (114). Injecting the pellet (116) through a pellet injection tube (108) at some distance from the toroidal field coils (110) allows the pellet to have no field in its interior when it enters the magnetic cage of the magnetic fusion device. Once the pellet (116) is inside the magnetic cage the field is initially excluded from the pellet's interior by circulating eddy currents on the pellet's surface. These eddy currents, by Lenz's law, oppose the diffusion of the magnetic field into the pellet for some amount of time. The soaking in time or magnetic diffusion time depends on the electrical conductivity of the pellet. In some embodiments, the magnetic fusion device includes cryostat storage device (102) for storing the pellets (116). The cryostat storage device (102) is connected to the pellet injector (106). In some embodiments, the cryostat storage device (102) can be connected to the pellet injector (106) by a pellet transfer device (104). By using the cryostat storage device (102), the electrical conductivity of the pellet is increased by pre-cooling the pellet to ultra-low temperature. In some embodiments, the pellets may be cooled to near absolute zero by immersing the pellets in a liquid helium bath. In some embodiments, the pellets may be cooled in a liquid helium bath to less than or equal to 40 kelvin (K). In an exemplary embodiment, the pellets may be cooled in a liquid helium bath to approximately 10 kelvin (K). FIG. 1 shows an exemplary cross section of a magnetic fusion device (100) showing the plasma (111) with nested magnetic flux surfaces (112) which confine the plasma, and the location of the toroidal magnetic field coils (110) which produce the strong magnetic field inside the plasma (coming out of the page), with zero magnetic field outside the toroidal coil cage. In some embodiments, the pellets (116) are stored in a cryostat storage device (102) and cooled to 40 K before they are injected into the plasma (111) at high velocity V. Since the pellets (116) are launched in a field-free region the pellets have no magnetic field inside. When the pellets exit the pellet injection tube (108), they are suddenly exposed to the toroidal magnetic field, which is suddenly screened from the interior of the pellet by the diamagnetic effect. As the pellet (116) transits the plasma (111) the field is screened out by the diamagnetic currents flowing on the surface of the pellet, thereby reducing the plasma heat flux falling on the surface of the pellet. The characteristic time for diffusion of the field into the pellet τd is comparable to the time τp for the pellet to reach the central region of the plasma that can be about 4 ms for a pellet injected into the magnetic fusion device at a velocity of approximately 500 m/s. Without magnetic shielding the pellet may completely vaporize before reaching the plasma core. The pellet still has a small residual resistivity, so that the field diffuses into the pellet on the characteristic time scale τd set by the residual. FIGS. 4A-4D show simulations of the magnetic configuration surrounding an exemplary ablating light metal pellet cooled to 40 K, where t′=t/τ, τ=μ0σrp2/π2, rp is the pellet radius, and σ is the electrical conductivity of the pellet. The simulations shown in FIGS. 4A-4D use spherical coordinate system (r,ϑ,ϕ) where the center of the pellet (400) is located at the origin r=0. The pellet (400) is normalized to a unit circle r=1. FIGS. 4A-4D show actually projections of the fields and the flows on a constant ϕ plane cutting through the symmetry axis ϑ=0 which in the figure corresponds to the x-axis, so that x=r cos ϑ and y=r sin ϑ. The solid lines drawn from the left to the right of each figure denote selected magnetic streamlines. The annular region between the surface of the pellet (400) on the unit circle, and the first dotted line (402) is the neutral gas part of the ablation flow. The second dotted line (404) illustrates the surface where there is a shock, transitioning the purely radial flow along the r direction to parallel flow along the distorted magnetic field lines. The region between the first dotted line (402) and the second dotted line (404) is the ionized ablation flow region. In FIG. 4A, the arrows show ablation flow vectors near ablating pellet. Diffusion of field into the pellet, known as eddy current decay, is incomplete because of outward ionized ablation flow. The field at t=0 is initially completely excluded from the pellet, initially reducing the heat flux to zero. For t>0, the magnetic field diffuses into the pellet while the pellet is moving through the plasma. Due to the outward flow of ablated and ionized gas the magnetic field inside the pellet can be still reduced even for long times. Further analysis includes the time evolution of the magnetic field structure near the pellet, assuming that the currents and magnetic fields remain axisymmetric, where the axis of symmetry is in the direction of the straight undisturbed magnetic field far away from the pellet. In this exemplary model, the magnetic screening effect is enhanced by the outward expansion of the ablated and ionized gas. The ablation outflow is relatively undisturbed near the pellet where its pressure is considerably larger than the magnetic pressure. Thus, near the pellet, the outflow is nearly spherically symmetric and can flow across the magnetic field. A flow of ionized gas across a magnetic field induces an electromagnetic field (EMF), which by Ohms law creates an azimuthal current within the ionized gas. This current flows in the same direction as the eddy current flowing in the pellet. The two currents add, thereby enhancing the screening effect in the immediate vicinity of the pellet. Due to expansion, the ablation pressure decreases with distance from the pellet, until at some distance the ablation pressure become comparable to the magnetic pressure and further expansion is arrested. Beyond that distance, there is little to no current flow as the expansion ceases and the ionized gas is forced to flow along the magnetic field. In some embodiments, the high electrical conductivity in the pellet and the finite conductivity of outward flowing ablation material can prolong the diffusion time of the magnetic field into the pellet. The exemplary model calculates the time for the magnetic field to soak into the pellet and finds that during the shielding period the ablation rate of the pellet can be significantly reduced, about 4-6 times in the magnetic fusion device plasmas. As shown in FIGS. 4A-4D, the magnetic field strength is greatly reduced near the surface of the pellet along with the heat flux which is directly proportional to the intensity of the field. FIG. 5 illustrates an exemplary method (500) of adding the exemplary pellets in a magnetic fusion device. The exemplary method comprises a confining operation (502) where a magnetic cage of is used to confine plasma in a plasma vessel. In the storing operation (504), the exemplary pellets are stored in a pellet storage device. In an optional transferring operation (506), a pellet transfer device can transfer the pellets from the pellet storage device to a pellet injector. In the injecting operation (508), a pellet injector can inject the pellets into the plasma vessel to expose the pellets to the plasma. In some embodiments, a cooling operation may also be performed prior to the transferring operation (506). The cooling operation may include using a cryostat storage device to cool the pellets. The exemplary embodiments pertain to methods for enhancing the penetration depth of a projectile injected into hot plasma in a plasma confinement device, for example, a tokamak plasma device. The injection of impurity particles deep in the plasma may protect the plasma confinement device against severity of plasma disruption events. In some embodiments, the disclosed technology uses the injection of metallic impurity pellets to achieve a controlled plasma shutdown by radiating away the plasma thermal energy over a large area. The disclosed technology also enables the capacity for deep penetration by engaging the phenomena of magnetic shielding. The magnetic shielding refers to the strong diamagnetic property of metallic materials when cooled to cryogenically low temperatures where the electrical resistance becomes extremely small, thereby increasing the electrical conductivity. The diamagnetic property is not permanent. However the duration of the effect of the diamagnetic property at low temperatures is such that the confining magnetic field which confines the plasma will be excluded from the interior of the pellet during its transit through the plasma, which in turn temporarily shields the pellet from the intense plasma heat flux. FIG. 6 shows a cross section of another exemplary pellet, such as a shell pellet that may be used in the magnetic fusion device without cryogenic treatment. The exemplary shell pellets can be stored in a storage device and can be sent to the pellet injector using a pellet transfer device that sends the exemplary pellets to the plasma vessel. The exemplary shell pellet (600) includes an outer shell (602) and the payload (604) that may include densely packed granules or a porous material. In some embodiments, a thin-shelled hollow pellet (600) is filled with payload (604) such as small granules in the size range of 10 to 200 microns. The small granules may be of a metallic or insulating material, which may serve as the payload for the shell pellet. The shell (602) undergoes ablation while in flight through the hot plasma while shielding the interior payload from the plasma heat flux until at some specified point deeper inside the plasma the shell (602) disintegrates exposing the payload (604) to the plasma. Once that happens, the particles begin to heat up and melt or vaporize. In the molten state, the granules may undergo fragmentation into smaller droplets due to the shearing action of a non-uniform surface ablation pressure. Solid pellets do not break up because the shear strength of the pellet material usually exceeds the ablation pressure non-uniformity. As fragmentation into smaller particles greatly increases the surface to volume ratio, both frictional deceleration and ablation (mass loss) rates dramatically increase. As a result the particle payload substance can be more easily dispersed and trapped inside a radiatively cooled plasma, enabling plasma densification. In some embodiments, the shell (602) encapsulating the payload (604) may include a low-Z material of a desired or adequate structural strength to withstand acceleration forces in the pellet injection tube without breaking up. The materials for the exemplary shell (602) may include, for example, light metals such as lithium (Li), or beryllium (Be), or related insulating compounds, such as lithium deuteride (LiD), or beryllium deuteride (BeD), or boron nitride (BN). In some embodiments, the payload (604) inside shells may include similar materials as their shells, including, for example, lithium (Li), lithium deuteride (LiD), beryllium (Be), beryllium deuteride (BeD), boron (B), boron nitride (BN), or tungsten (W). The chemical compositions for the cooled pellets and non-cooled pellets can be designed or chosen to reduce or minimize contaminates into the plasma vessel. For example, lithium may be used on the inner walls of some plasma vessels to improve overall plasma performance. The exemplary shells and payloads of the exemplary pellets for such plasma vessels can include lithium that can mitigate plasma disruption and can minimize introducing contaminates into the plasma vessel. Similarly, for non-cooled pellets, the use of the light metals, such as lithium can also minimize unwanted elements inside the vacuum vessel. A benefit of using light metal low-Z pellets is that low-Z pellets can be cryogenically cooled to engage the magnetic screening effect as described in this patent document. A benefit of using pellets comprising insulating compounds is that they allow for a higher heat of vaporization and heat of dissociation for the molecular compounds which reduces the ablation rate and promotes deeper penetration. Another benefit of using low-Z shells is that such use allows the plasma to radiatively cool at a slower rate. Resistive Magnetohydrodynamics (MHD) theory warns to be careful not to cool the plasma on the q=2 magnetic resonance surface by lingering impurities. Otherwise, a resistive kink or tearing mode can destroy the outer magnetic surfaces and trigger an inward propagating cooling front. Accordingly, cooling from the inside out is preferred over cooling from the outside in because the central region of the plasma has fewer rational or unstable magnetic surfaces than the outer regions. The ablated and ionized particles cause the plasma to undergo a radiatively driven thermal collapse in the deeper regions of the plasma. The particulates can also dissipate runaway electrons, stunting the seed runaways that can amplify by the avalanche knock-on effect into large and dangerous runaway electron currents. Based on simulations, when the plasma is cooled from fusion temperatures (e.g., approximately 20,000 eV) down to lower temperatures (e.g., approximately 30-100 eV), the initially densely packed particles may still rapidly break up into smaller clusters which then recursively fission into smaller and smaller ones, until individual dust particles eventually become exposed to the plasma. An estimate of the disassembly time can be given as follows. The time to cause the first fission of the initial payload cluster is roughly τf0=R0/c0 where R0 is the initial pellet radius and c0=α√{square root over (p0/ρ)} is a sound time with p0 being the surface ablation pressure exerted over the initial payload cluster ρ is the dust mass density including the void space and α<1 is an asymmetry coefficient. After n fissions the number of densely packed particle clusters or blobs will be 2″. It is further assumed that: (1) mass is conserved (neglect blob ablation); (2) each blob is spherical in shape; and (3) ρ and α for all of the blobs remains constant. Given that the ablation pressure scales with the blob radius as ρ∝R−1/3, the time for the original payload to disperse in this way becomes τ frag = τ 0 ∑ n = 0 ∞ ( 1 2 ) 7 n / 18 = τ 0 1 - ( 1 / 2 ) 7 / 18 = 4.2322 τ 0 Taking the initial payload cluster to be the diameter of the pellet, for example, D=1.5 cm, and ρ˜1 g/cm3, the ablation pressure in a plasma cooled to 100 eV plasma is approximately 1 MPa, giving the fragmentation time of about 1 ms, which is sufficiently short if the pellet speed is <500 m/s. Penetration of dispersed individual dust particles will be limited by both deceleration and mass loss (ablation). The faster vaporization rate is due to the extremely small particle size; the lifetime being proportional to τlife∝D3/G∝D5/3 (where D and G are the grain diameter and mass ablation rate, respectively) assures that isolated granules will completely ablate inside a warm 30-100 eV plasma rather than hitting the opposite wall. The particulates are also contained inside the plasma by friction stopping. In some embodiment, the size of the fine grains or dust may be approximately 10 microns, much less than the Debye length of the ambient plasma. Then the Coulombic drag force on an isolated dust grain can be calculated from the orbital motion limited (OML) treatment. Assuming that the initial dust particle velocity relative to the plasma is equal to the original pellet velocity Vpellet approximately 500 m/s. Since Vpellet is much less than the thermal speed of the ions and electrons even for a 10-100 eV thermally collapsed plasma, the Coulombic drag force on a moving dust particle is largely due to its interaction with the background plasma ions. The stopping time r of a dust grain moving relative to the background ions is therefore analogous to the slowing down time of a fast (MeV) ion test particle moving through a hot electron plasma, in the sense that the scaling is similar: τ∝T3/2/Zd2, where T is the temperature of the target particles and Zd is the charge on the dust grain (test particle). The 3/2 temperature scaling is familiar, and it indicates that the stopping time shortens as the plasma radiatively cools, except in the case of a dust grain it adopts a potential, the floating potential, so that its charge Zd also depends on plasma temperature T. Putting this all together, a slowing down time in a hydrogenic plasma is shown below: τ = 7.685 × 10 22 ρ d r n ln Λ ( MT ) 1 / 2 ψ 2 where n(m−3), T(eV), and M(amu) is the plasma ion density, temperature, and ionic mass, respectively, ρd(kg/m3) is the mass density of the material in the dust grains, r(m) is the grain radius, and ψ˜2-5 is the floating potential normalized to the ion temperature, and ln Λ approximately 10 is the Coulomb logarithm. As an example, taking fusion parameters such as T=100 eV and n=1020 m−3, we get τ=0.6 ms, so in some embodiments, a stopping distance for pellet speed of approximately 500 m/s may be 30 cm. Therefore, any unablated dust particles can be easily trapped inside plasma with minor radius of 0.66 m. While this patent document contains many specifics, these should not be construed as limitations on the scope of any invention or of what may be claimed, but rather as descriptions of features that may be specific to particular embodiments of particular inventions. Certain features that are described in this patent document in the context of separate embodiments can also be implemented in combination in a single embodiment. Conversely, various features that are described in the context of a single embodiment can also be implemented in multiple embodiments separately or in any suitable subcombination. Moreover, although features may be described above as acting in certain combinations and even initially claimed as such, one or more features from a claimed combination can in some cases be excised from the combination, and the claimed combination may be directed to a subcombination or variation of a subcombination. Similarly, while operations are depicted in the drawings in a particular order, this should not be understood as requiring that such operations be performed in the particular order shown or in sequential order, or that all illustrated operations be performed, to achieve desirable results. Moreover, the separation of various system components in the embodiments described in this patent document should not be understood as requiring such separation in all embodiments. Only a few implementations and examples are described and other implementations, enhancements and variations can be made based on what is described and illustrated in this patent document. |
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054003750 | summary | BACKGROUND OF THE INVENTION The present invention relates to a technology for transmuting transuranium elements and more particularly to a transuranium transmuting reactor core for transmuting the transuranium elements at a fast reactor and also to a transuranium elements transmuting fuel pin and fuel assembly charged into a reactor core of a fast reactor. A spent fuel discharged from a thermal reactor such as boiling water reactor or the like includes transuranium elements (hereinafter called TRU elements) such as neptinium-237 (.sup.237 Np), americium-241 (.sup.241 Am), americium-243 (.sup.243 Am), curium-242 (.sup.242 Cm), curium-244 (.sup.244 Cm) and others which are high-level radioactive wastes, and in minor actinides (hereinafter called MA elements) present after eliminating plutonium (Pu) from the TRU elements, there exists elements such as .sup.237 Np, .sup.241 Am, .sup.243 Am or the like having an extremely long half life such as 2.14 million years, 432 years, 7,380 years, which cannot be quenched within a short period of time. Thus, it is desired that the MA elements are transformed into elements with a short half life through a nuclear transmutation in a short period of time. A prior art includes technique for transmuting the TRU element which comprises using a fast reactor extremely high in a neutron energy as compared with a thermal reactor and subjecting the TRU elements charged into a fuel charged in a core of the fast reactor to a nuclear transmutation ((1) "Conceptional Design Study on Actinide burning Fast Reactor", T. Osugi et al., JAER1-M 83-217, issued by Japan Atomic Energy Research Institute in December 1983; (2) "Transmutation of Transuranics in FBR", A. Sasahara, T. Matsumura, F7, Fall Meeting Reports, Atomic Energy Society of Japan, 1988). The prior art TRU elements transmuting comprises transmuting the aforementioned MA elements by causing a transmutation shown in FIGS. 9A to 9C to the typical MA elements of .sup.237 Np, .sup.241 Am and .sup.243 Am which are main objects of transmuting at a fast reactor core. In FIGS. 9A to 9C, F.P. denotes fission products, and elements given in a square border around indicates that of being easy to cause a fission against a neutron energy in the fast reactor, namely, that its energy averaged fission cross-sections are about 1 burn or over. The prior art TRU elements transmuting process utilizes a feature of the fast reactor core effectively, and the feature comes in: (1) Since a neutron energy of the fast reactor core is high, a neutron capture is hard to occur in .sup.237 Np, .sup.241 Am and .sup.243 Am and the like, and thus an evil influence of the fast reactor on a neutron economy according to the charging of the TRU elements into the reactor core is relatively small (a neutron capture cross-section getting small according as the neutron energy becomes high as shown in FIG. 18). PA1 (2) The fast reactor is-generally high by about 1 digit in a neutron flux level as compared with the thermal reactor, therefore the TRU elements can be subjected to a nuclear transformation even if a fission and neutron capture cross section on an energy average is small, and thus a high transmuting efficiency of the TRU elements is ensured. PA1 (1) If the MA elements to be transmuted is added to uranium-plutonium mixed fuel, a melting point of the mixed fuel lowers. Then the melting point drop is capable of causing a fuel melting, thus a measure such as lowering a reactor power or the like will be necessary for avoiding the fuel melting, which may deteriorate the transmuting efficiency of the MA elements. PA1 (2) As will be apparent from FIGS. 9A to 9D, the typical MA elements to be transmuted is generally hard to bring about a fission, and hence is transformed into fissionable elements by a neutron capture. Accordingly, if the fast reactor core is charged with the TRU element excessively much, then, as shown in FIG. 19, an amount of fissionable elements produced newly by the neutron capture of the MA element according to a neutron irradiation comes to exceed fissionable elements transmuted by fission, thus an excess reactivity of the fast reactor increasing. PA1 (3) The TRU elements to be transmuted are easy to cause an alpha-decay in most cases, and an alpha ray energy emitted at the time of the alpha-decay is relatively high at 4 to 6 MeV generally. Accordingly, if the MA elements are added much to a fuel, a calorific value and a source intensity of gamma ray, neutron and others become excessive from the state of a fresh fuel before loading into the fast reactor core. Further, at the time of assembling, storage and transportation of new fuel assemblies in which the MA elements are enclosed, a heat removing of the alpha ray energy becomes difficult and the fuel overheats to lead to a failure in a worst case. PA1 (4) When charging a fast reactor uniformly with the TRU elements to be transmuted at the core with a core for which a plutonium enrichment is one kind as a base, a radial distribution of the power density, namely a radial power distribution during operation of the reactor becomes small according as it comes outside, as shown in FIG. 20, therefore a transmuting efficiency of the TRU element and a plant power generation efficiency being unsatisfying. PA1 (5) When charging the reactor uniformly with the TRU elements at the core with the fast reactor core for which a plutonium enrichment is two or more than two kinds as a base, a radial power distribution of the core is improved as compared with FIG. 20 by an adjustment of the plutonium enrichment, a flatting requirement can thus be satisfied, however, as shown in FIG. 21, for example, there arises a portion where the power distribution largely fluctuates according to burn-up. In the prior art transmuting of the TRU elements, nothing has been taken particularly into consideration for charging amount of the TRU elements charged into a fast reactor core and its distribution in core when carrying out a transmuting of the TRU elements. Still, however, only a self-evident technical care on charging the core with the TRU elements as much as possible has been considered for enhancing a transmuting efficiency of the TRU elements. However, if the fast reactor core is charged with the TRU elements as much as possible, then the following problems are capable of resulting therefrom. Consequently, if the charging amount of the TRU elements and its distribution are not specified properly, an excessive change or distortion may arise on a reactor power distribution and a neutron flux distribution, thus leading to problems on safety and characteristics of the reactor. On the other hand, a flow rate of a coolant flowed for cooling down the fast reactor core is constant through the lifetime of a reactor plant. The flow rate of the coolant to fuel assemblies is set adaptively to the time when the power is maximized. Thus, when the output distribution fluctuates largely according to the burn-up of the fuel, a heat removing efficiency deteriorates, a heating efficiency gets lowered furthermore, which is not preferable from the viewpoint of an economical operation of the reactor plant. SUMMARY OF THE INVENTION An object of the present invention is to substantially eliminate defects or drawbacks encountered in the prior art described above and to provide a transuranium elements transmuting reactor core capable of transmuting the TRU elements efficiently without causing a failure of the fuel assemblies, increase of excess reactivity, deterioration of thermal efficiency and others. Another object of the present invention is to provide a transuranium element transmuting fuel and fuel assembly capable of preventing the lowering of the power density of a fast reactor and the distortion of the power distribution of the fast reactor and effectively transmuting the TRU elements. These and other objects can be achieved according to the present invention by providing, in one aspect, a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of a transuranium element to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting, and in the improvement, the amount of the transuranium elements to be added to the fuel assemblies is controlled so as to keep an excess reactivity of the reactor substantially zero through operation of the reactor. In another aspect, there is provided a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of a transuranium element to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting, and in the improvement, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am are set so as to satisfy an equation EQU 1.2.times.10.sup.2 .times.M.sub.242 +2.8.times.M.sub.244 +1.1.times.10.sup.-1 .times.M.sub.241 <Q.sub.1 where an upper bound of heating rates of the single fuel assembly outside the reactor is .sub.1 Q from the view point of the fuel assembly integrity, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am and also satisfy an equation EQU 1.2.times.10.sup.2 .times.M.sub.242.sup.L +2.8.times.M.sub.244.sup.L +1.1.times.10.sup.-1 .times.M.sub.241.sup.L <Q.sub.2 where an upper bound of the heating rates, per unit length of the fuel pellet contained in the fuel pins is Q.sub.2 from the view point of the fuel element integrity, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am per the unit length are M.sub.242.sup.L, M.sub.244.sup.L and M.sub.241.sup.L. In a further aspect, there is provided a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of transuranium elements to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting and in the improvement, a charging density of minor actinides is set to lessen outwards of a core central portion in a core area where a plutonium content is made even. In a still further aspect, there is provided a transuranium element transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of transuranium elements to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting and in the improvement, a charging density of minor actinides is set high accordingly in an area where a plutonium is enriched high at the core of a plutonium enriched area where a plutonium content varies. In a still further aspect, there is provided a transuranium element transmuting fuel pin wherein a transuranium fuel pin is formed by charging a transuranium fuel material in a fuel clad and the transuranium fuel material includes at least one of fuel materials consisting of an enriched uranium and a uranium-plutonium mixed fuel and a fertile material consisting of a degraded uranium, a natural uranium and a depleted uranium contain transuranium elements such as Np, Am and Cm. In a still further aspect, there is provided a transuranium element transmuting fuel assembly including a wrapper tube and a plurality of fuel pins enclosed in the wrapper tube, each of the fuel pins including a fuel clad, wherein at least one part of the fuel pins are formed by charging a transuranium fuel material in the fuel clad with a transuranium fuel material inside. In a preferred embodiment, the fuel pins enclosed in the wrapper tube comprises transuranium fuel pins charged with the transuranium fuel material and fuel material pins charged with a fuel material consisting of an enriched uranium and a uranium-plutonium mixture fuel, and a radioactive fission product such as Sr or alkaline metals is contained in the transuranium fuel material. In the transuranium element transmuting reactor core according to the present invention, since an amount of transuranium elements to be added to a fuel pin of the fuel assemblies is controlled so as to keep an excess reactivity of the reactor substantially zero through an operation of the reactor, a decrease of effective multiplication factor according to the lapse of time for operation will be prevented, an excessive deterioration or turbulence of the reactor power distribution can be prevented, and as looking for improvement of a power plant capacity factor from enhancing a reliability of the plant, transuranium elements (TRU elements) can be transmuted efficiently. Further, from setting loading amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am so as to realize: EQU 1.2.times.10.sup.2 .times.M.sub.242 +2.8.times.M.sub.244 +1.1.times.10.sup.-1 .times.M.sub.241 <Q.sub.1 where an upper bound of the single fuel assembly power assembly outside the reactor is .sub.1 Q from the view point of the fuel assembly integrity, loading amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am which can be loaded into the single fuel assembly are M.sub.242, M.sub.244 and M.sub.241, and also to realize: EQU 1.2.times.10.sup.2 .times.M.sub.242.sup.L +2.8.times.M.sub.244.sup.L +1.1.times.10.sup.-1 .times.M.sub.241.sup.L <Q.sub.2 where an upper bound of the heating per unit length of the fuel element contained in the fuel assemblies is Q.sub.2 from the view point of the fuel element integrity, charging amounts of .sup.242 Cm, .sup.244 Cm and .sup.241 Am per the unit length are M.sub.242.sup.L, M.sub.244.sup.L and M.sub.241.sup.L, a melting of the fuel element during operation of the reactor and an overheating or failure of the fuel assemblies outside the reactor can effectively be prevented, and an accident of a control rod and a neutron absorbing material of the control rod can be reduced by a neutron absorption effect of .sup.242 Cm, .sup.244 Cm and .sup.241 Am, an enhancement of heat removing efficiency of the core can thus be realized, an economical operativity is also improved, and a safety and reliability of the core and the fuel assemblies are ensured as well, thus transmuting the TRU elements efficiently. Further, by setting a charging density of minor actinides to lessen outwards of a core center in a core area where a plutonium content is even, and also by setting a charging density of minor actinides high accordingly in an area where Pu is enriched high at the core of a Pu-enriched area where a plutonium content varies, a flatting requirement of a radial distribution of the reactor power can be satisfied, an enhancement of safety and reliability of the core and the fuel assemblies will be realized without causing the excessive deterioration and turbulence of the reactor power distribution, thus traansmuting the TRU elements efficiently. In a further aspect, according to the transuranium element transmuting fuel assembly of the characters described above, even if the transuranium fuel material is charged in the transuranium fuel pin, the degradation of the core power density and the distortion of the core axial power distribution can be effectively prevented, thus improving the core cooling efficiency and effectively transmuting the transuranium element. |
050193252 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT General Overview of the Structure and Operation of the Invention With reference to FIG. 1, wherein like numerals designate like components throughout all these several figures, the control rod drive removal and installation system 1 of the invention generally comprises an installation and removal assembly 2 that includes a carriage 3 movable along the service rails in the undervessel, a cradle 4 pivotally connected to the carriage 3, in combination with a modified transfer cart 5 which is likewise movable along these service rails. The transfer cart 5 advantageously includes a raising and lowering mechanism 6 for raising and lowering control rod drives from the underside of the cradle 4 when this cradle is pivoted into the horizontal position illustrated in FIG. 2B. The principal purpose of the system 1 is to remove and install a control rod drive 7 from the tubular drive housings 9 extending down from the bottom of the reactor vessel (not shown) of a boiling water reactor (BWR). These control rod drives 7 extend down into the top portion of the undervessel cavity 11 between an array of pipe-like instrument tubes 13 and a "forest" of electrical cables (not shown). The sides of the undervessel cavity 11 are defined by a cylindrical wall structure 15 having an access way 17. A pair of service rails 19 traverses the diameter of the cylindrical wall structure 15 defining the undervessel cavity 11. These service rails 19 are flanked by service platforms 20a, 20b having guardrails 21 which provide maintenance personnel with a surface to stand upon when manipulating the carriage 3 or the transfer cart 5 of the system 1. The service rails 19 are formed from a pair of parallel channel beams whose upper and lower flanges define upper tracks 22 and lower tracks 23. As may best be seen with respect to FIG. 2D, an inverted angle iron 25 is welded along one of the upper tracks 22. Additionally, a guide rail 27 having a capturing flange 29 is welded over the upper track defined by the angle iron 25. The combination of the upper V-shaped surface provided by the angle iron 25, and the capturing flange 29 of the guide rail 27 serves to secure the grooved wheels 30a of the carriage 3 to the upper tracks 22 so that there is no danger of the installation and removal assembly 2 from rolling off the upper tracks 22 during operation. The ends of the channel beams forming the service rails 19 are rollably mounted upon a circumferential rail 31 which completely surrounds the inner surface of the cylindrical wall structure 15. The circumferential rail 31 allows the service rails 19 to be rotated 360 degrees around the cylindrical wall structure 15 by means of an electrically-operated drive train (not shown). The 360 degree movement afforded by the rail 31, in combination with the radial movement afforded by the service rails 19, allows the carriage 3 and the modified transfer cart 5 to be moved into virtually any selected point under the reactor vessel in polar-coordinate fashion. To guide both the carriage 3 of the installation and removal assembly 2 and the transfer cart 5 onto the service rails 19, transition rails 33 are provided through at least one of the access ways 17 leading into the cylindrical wall structure 15. While the cylindrical wall structure 15 is not drawn entirely to scale in FIG. 1, it can still be appreciated that there is very little diametral or vertical clearance afforded within the undervessel cavity 11. With specific reference now to FIGS. 1, 2A and 2B, the cradle 4 of the installation and removal assembly 2 is connected to the carriage 3 by means of a pivot joint 39. The cradle 4 is pivotable from the vertical position illustrated in FIG. 1 to the horizontal position illustrated in FIG. 2B by means of pivot drive 41, which includes a single hydraulic cylinder 43. In the preferred embodiment, hydraulic cylinder 43 is a single action cylinder having a three and one half inch bore and a piston rod that is approximately two inches in diameter and having a forty inch stroke. Such a cylinder 43 is available from Hydra-Dynamics Corporation located in Kenilworth, Ill. As is best seen in FIG. 1, the hydraulic cylinder 43 is connected between the back end of the carriage 3 and the upper end of the cradle 4. Such connection points allow the hydraulic cylinder 43 to exploit the small amount of mechanical advantage that is provided in the length L1 of the cradle 4 disposed between the pivot joint 39 and the upper end of the cradle 4 (see FIG. 2B). To render the installation and removal assembly 2 as compact as possible, the carriage 3 is formed from a pair of side panels 46a, b which define a hollow, rectangular interior 60 that is complementary in shape to the cradle 4, and which allows the cradle 4 to be folded completely within the cradle interior 60 when it is disposed in the horizontal position illustrated in FIG. 2B. A drive train 48 rotates the wheels 30a, b along the upper track 22. As will be discussed in detail later, the drive train incorporates a hydraulic motor which precisely positions the cradle 3 at a selected point along the service rails 19. The cradle 4 includes, on its underside, a retaining bracket 50 capable of capturing and retaining the conical end of a control rod drive 7. The retaining bracket 50 contains a centrally disposed opening 51 for a purpose which will be described later. The cradle 4 further includes an extendable and retractable jaw assembly 52 capable of selectively capturing the tubular body of a control rod drive. As will become more evident hereinafter, the jaws of the jaw assembly 52 do not clampingly grasp the body of the control rod drive; they merely circumscribe the drive 7 so that it may be freely and slidably moved upwardly and downwardly by the lifting and lowering mechanism 54 also included within the cradle 4. Together, the retaining bracket 50 and the extendable and retractable jaw assembly 52 provide a two point securing means for securing a control rod drive 7 during a lifting or lowering operation. The grasp of the jaw assembly 52 is further designed to be compliant to facilitate alignment between a control rod drive 7 and a drive housing 8. The lifting and lowering mechanism 54 contained within the cradle 4 consist of two components, including a lead screw drive assembly 56 which is threadedly engaged to the retaining bracket 50 and a two stage hydraulic cylinder 58 having a retaining cup 59 at its distal end for capturing the conical end of a control rod drive 7. As will be described in more detail hereinafter, lead screw drive assembly 56 is formed from a pair of parallel, threaded rods rotatably mounted within the cradle 37 and powered by a hydraulic motor. The clockwise or counterclockwise rotation of the parallel lead screws of the lead screw drive assembly 56 moves the retaining bracket 50 between the top and bottom ends of the cradle 4. However, as a control rod drive 7 must be lifted considerably beyond the upper end of the cradle 4 in order to become installed within a drive housing 8, the two stage hydraulic cylinder 58 is provided. This cylinder 58 is movable from a position within the cradle interior 60 to a position directly beneath the retaining bracket 50 when the bracket 50 is moved to the top end of the cradle 4 (as is shown in FIG. 1) by means of the combination of a toggle linkage 62 which is swung from an inner to an outer position by means of a hydraulic kickout cylinder 64. When the two stage hydraulic cylinder 58 is disposed beneath the retaining bracket 50 as is shown in FIG. 1, the retaining cup 59 may freely extend through the opening 51 in the retaining bracket 50 to complete the lifting of a control rod drive 7 into an installed position within a drive housing 8 (as is indicated in phantom in FIG. 1). With reference again to FIG. 2B, the length L of the cradle 4 is deliberately made to be approximately twelve feet, which is about three feet shorter than the fifteen foot length of a control rod drive 7. Such a shorter length advantageously reduces the overall length of the assembly 1, thereby rendering it more compact and more easily maneuverable through the cavity access ways 17 and into the undervessel cavity 11 itself. The shorter length also advantageously reduces the overall weight of the assembly 1. Further complementing this design feature is the fact that the length L3 of the carriage 3 is considerably shorter than the length L of the cradle 4, and that the pivot joint 39 is located along the longitudinal axis of the cradle 4 at a point that the entire length L3 of the carriage will be subsumed within the length L of the cradle when the cradle 4 is positioned horizontally. A final dimensioning feature which advantageously assists the assembly 1 in performing its operation without mechanical interference and without the necessity for constantly repositioning the carriage 3 when the cradle 4 is pivoted upwardly results from the fact that the pivot joint 39 is located along the longitudinal axis of the cradle 4 at a point which allows only a short upper length L1 to extend upwardly toward the instrument tubes 13 and hydraulic lines (no shown) when the cradle 4 is pivoted into the vertical position illustrated in FIG. 1. Such proportioning has the undesirable effect of creating a mechanical disadvantage in the pivoting movement of the cradle 4 which requires a considerable amount of drive force from the pivot drive 41, as the upward length L1 of the cradle 4 is considerably smaller than its downward length L2. However, this mechanical disadvantage is overcome by the use of the previously described hydraulic cylinder 43, which is not only amply powerful enough to reliably and smoothly effect the pivoting motion, but which further is capable of producing the required power with only a minimal weight addition to the overall assembly 1. Moreover, because of the aforementioned mechanical disadvantage which occurs as a result of the fact that L1 is much smaller than L2, the cylinder 43 need only provide a relatively short stroke in order to effect the desired pivoting movement. The proximity of the cylinder 43 to the carriage 2, coupled with the relatively short, horizontal stroke that it provides to effect the desired 90 degree pivoting motion, helps to provide a tooling system 1 with short linear dimensions and compact operational strokes and movements which helps to avoid mechanical interference between the assembly 2 and the control rod drives 7, low-hanging instrument tubes 13 and electrical cables extending down from the bottom of the reactor vessel. Specific Description of the Structure and Operation of the Invention With reference again to FIGS. 2A and 2B, the drive train assembly 48 of the carriage 3 includes a hydraulic motor 70 whose output shaft is coupled to a drive sprocket 73. The use of a hydraulic motor to power the drive train assembly 48 is preferred for two reasons. First, the rpm output of such a hydraulic motor may be easily and accurately controlled by merely monitoring the volume of pressurized fluid conducted through it, which in turn allows the maintenance personnel to remotely move the cradle 3 to a precise point along the longitudinal axis of the service rails 19. Secondly, unlike electrical motors, hydraulic motors are uneffected by the water which often splashes down from the drive housings 8 when the worn out control drives 7 are removed for servicing. In the preferred embodiment, a Char-lynn hydraulic motor manufactured by the Eaton Fluid Power Division of Eaton Corporation located in Eden Prarie, Minn., is used to power the drive train assembly 48. The previously mentioned drive sprocket 73 is engaged to driven sprockets 75 and 77 connected to the two, forward-most wheels 30a of the carriage 3. The engagement of the drive chain 74 to two of the grooved wheels 30a ensures that, at all times, a positive traction will exist between the carriage wheels 30b and the angle iron 25 protruding up from one of the upper tracks 22. Such positive traction allows the equipment operator to determine the precise position of the assembly 2 along the service rails 19 by monitoring the number of turns of the output shaft of the motor 70. The driven sprockets 75 and 77 are larger than the drive sprocket 73 to provide a three to one drive ratio. An idler gear 79 is provided in order to adjust the tautness of the drive chain 74. With reference to FIGS. 2A. 2B and 2C, the cradle 4 is formed from a pair of aluminum channel beams 66a, 66b tied together along their top sides by a reinforcing plate 67. The interior 60 defined by the opposing beams 66a, 66b contains both the lead screw drive assembly 56 and the two stage hydraulic cylinder 58 of the lifting and lowering mechanism 54. The back end of the cradle 4 further includes a rear pivot bracket 82 comprised of a pair of parallel bracket plates 83a and 83b whose bottom edges are integrally connected to a reinforcing plate 84 which in turn is welded to the side panels 46a and 46b of the carriage 3. A yoke-type pivot pin 85 extending through bores in the bracket plates 83a, 83b links these plates to the rear lug 87 of the pivot cylinder 43. This reinforcing plate 84, along with spacer rib 89 (located near the middle of the carriage 3) secures the two side panels 46a and 46b together in a parallel relationship while still leaving a rectangular space 47 within the carriage 3 which the cradle 4 may be received into when the cradle 4 assumes a horizontal position. The front end of the cradle 4 includes a front pivot bracket 92 formed from a mounting lug 94 which is connected at its bottom portion to the top or front end of the cradle 4 and at its top portion to a piston rod clevis 96 by means of another yoke-type pivot pin 98. As is evident in FIG. 2B, the rear pivot bracket 82 and the front pivot bracket 92 mount the hydraulic cylinder 43 of the pivot drive 41 in a close, parallel relationship with respect to the cradle 4 and service rails 19. Such a configuration helps to localize all mechanical motions close to the body of the installation and removal assembly 3, and away from reactor components which could possibly interfere with the pivoting cradle 4. Moreover, making both of the brackets 82 and 92 approximately the same height with respect to the carriage 3 helps to minimize moment forces on the piston rod 44 when the cradle 4 is folded within the carriage 3 as shown in FIG. 2B by directing the load applied to the rod 44 into a load whose components are exclusively compressive in nature. With reference now to FIGS. 2B and 2C, the extendable and retractable jaw assembly 52 is located just in front of the previously described front pivot bracket 92. The purpose of this component is to detachably secure the tubular body of a control rod drive in such a way that it is easily slidable through the jaws along the longitudinal axis of the cradle 4. To this end, the jaw assembly 52 includes a jaw mechanism 103 formed from a pair of opposing jaws 105a and 105b pivotally linked together in "scissors" fashion by a pin 107 to a sliding member 108. A jaw-clamping, double-action hydraulic cylinder 109 is likewise connected to the slide member 108 by means of bracket 110. The clamping cylinder 109 has a reciprocable piston rod 111 linked to a clevis 112 which in turn is connected to a bracket 114 coupled to jaw 105a. In operation, the extension or retraction of the piston rod 111 causes jaw 105a to pivot toward or away from the jaw 105b. To permit a control rod drive 7 to slidably move along the axis of the cradle 4 through the jaw mechanism 103, each of the of the jaws 105a and 105b is provided with a roller 116a and 116b to permit the entire jaw mechanism 103 to be retracted from the positions illustrated in FIGS. 2B and 2C to a position where the mechanism 103 does not extend beneath the cradle 4, the slide member 108 is in turn connected to the piston (not shown) of a double action extension and retraction cylinder 122. The retracting movement afforded by the cylinder 122 allows the jaws 105a and 105b to be released from the body of a control rod drive 7 so that the rod retaining cup 59 located on the distal end of the piston rod of cylinder 58 will not come into contact with the jaw mechanism 103 when the cup 59 extends upwardly to lift a control rod drive 7 into an installed position in a drive housing 8. The extendable and retractable assembly 52, in combination with the retaining bracket 50 located on the underside of the carriage 4, provide a securing mechanism that secures a control rod drive 7 at two separate points along its longitudinal axis. In operation, the pressure of the hydraulic fluid used to power the hydraulic cylinders 109 and 122 is maintained at a level great enough to grippingly secure a control rod drive 7, but low enough to afford some compliance in this grip so that, when the control rod drive 7 is lifted up toward a drive housing 8, a maintenance worker can easily manipulate the distal end of the drive 7 into the open end of the housing 8. As is best seen in FIGS. 2B, and 2C, the lead screw drive assembly 56 of the lifting and lowering mechanism 54 includes a pair of parallel lead screws 125a and 126b that are rotatably mounted at their back and front ends within the cradle 4. Each of the front ends of the lead screw 125a and 125b terminates in a driven sprocket 127a and 127b respectively. These driven sprockets 127a and 127b are in turn driven via a drive chains 128a, 128b which mesh with the drive sprocket 129. Drive sprocket 129 is in turn connected to the output shaft of a hydraulic motor 131 of the same type as the hydraulic motor used to power the carriage drive train 48. The interface between the retaining bracket 50 of the securing means and the lifting and lowering mechanism 54 occurs at lead screw adaptor 133. This adaptor 133 includes two parallel ball nuts 134a, 134b (only one of which is shown) which ride independently upon threaded rods. Hence when the hydraulic motor 131 turns the drive sprocket 129 that rotates the lead screws 125a and 125b, the lead screw adaptor 133 will move the retaining bracket 50 either toward the back or the front end of the cradle 4. The use of a lead screw drive assembly 56 in the lifting and lowering mechanism 54 of the system 1 is advantageous for two reasons. First, such a lead screw-type mechanism allows the maintenance personnel to precisely and remotely control the position of the retaining bracket 50 along the longitudinal axis of the carriage 4 by merely monitoring the number of rotations that the hydraulic motor 131 turns the lead screws 125a and 125b. Secondly, it is mechanically reliable and not apt to breakage. Thirdly, because either of the lead screws 125a or 125b is strong enough to carry a control rod drive 7 by itself, and because the ball nuts of the adaptor 133 are not interconnected with one another, there is no danger of failure or collapse of the bracket 50 in the unlikely event that one of the lead screws 125a or 125b should break or otherwise fail. Instead, the remaining lead screw and ball nut would be capable to afford a controlled movement of the bracket 50 until appropriate repairs could be made. Turning next to the two stage hydraulic cylinder 58 which forms the second component of the lifting and lowering mechanism 54, this cylinder 58 includes a first cylinder 135 having a second cylinder 137 telescopically mounted therein. The second cylinder 137 in turn contains a piston rod 139. The long stroke afforded by the use of a two stage, hydraulic cylinder 58 allows it to move a control rod drive from a position substantially coterminous with the front end of the carriage 4 to an installed position within a drive housing 8 while at the same time allowing the entire cylinder 58 to be stored in the hollow interior 60 of the cradle 4 near its back portion when not in use. The cylinder 58 includes a rear lug 141 which is linked to the rear, parallel members 143 of the previously mentioned toggle linkage 62 by means of pivot pin 144. The front end of the piston 58 includes a front mounting yoke 145 which is linked to the front parallel members 147 of the toggle linkage 62 by means of pin 148. The rear parallel members 143 of the toggle linkage 62 are connected to the previously mentioned kickout cylinder 64 (which is a double action hydraulic cylinder) by means of a yoke bracket 151 which is pivotally connected thereto at its distal end. The proximal end of yoke bracket 151 is pivotally connected to a clevis 153 coupled to the distal end of the piston rod 154 of the cylinder 64. This clevis 153 is slidably connected to a guide rail 155. In operation, the kickout cylinder 164 retracts the two stage cylinder 58 within the interior 60 of the cradle 4 by retracting the piston rod 154 so as to swing the rear parallel members 143 toward the front of the cradle 4. To extend the two stage cylinder 58 to a position just under the retaining bracket 50 (which is done after the retaining bracket 50 moves a control rod drive 7 into a coterminous position with respect to the top end of cradle 4) the piston rod 154 of the kickout cylinder 64 is extended into the position illustrated in FIG. 2B, which not only extends the linkage 62, but advantageously allows the compressive load born by the two stage cylinder 58 to be born by the reinforcing plate 149. With reference now to FIGS. 3A and 3B, the transfer cart 5 of the system 1 includes a box-type frame 160 having front and rear cradle members 161a and 161b traversing its front and rear ends respectively. To uniformly distribute the load of a control rod drive laid across the cradle members 161a and 161b, three sets of wheels 162a, 162b and 162c are rotatably connected to axles 164a, 164b and 164c connected to the back, middle and front portions of the frame 160. Disposed throughout the middle portion of the cart frame 160 is a toggle or a parallelogram linkage 166. This linkage 166 is formed from rear parallel bars 168 pivotally connected to an axle 170, and front parallel bars 172 pivotally connected upon an axle 174. The parallel bars 168 and 172 are linked together by lateral parallel bars 176 by way of pins 178 and 180. A front and rear rollers 182 and 184 journalled upon shafts 183 and 185 are provided at the distal ends of the front and rear parallel bars 172 and 168, respectively. These rollers 182 and 184 facilitate movement of a control rod drive 7 to and from the front and rear cradle members 168a and 168b. A raising and lowering cylinder 187 is centrally mounted between the frame 160 and the toggle linkage 166 in order to raise and lower the rollers 182 and 184. This cylinder 187 includes a double action piston rod 189 that terminates in a clevis 191 pivotally mounted to the front parallel bars 182 as shown. This cylinder 187 further includes a rear lug member 193 that is in turn pivotally connected to shaft 195. When the piston rod 189 of the cylinder 187 is extended as is best seen in FIG. 3A, the rollers 182 and 184 are lifted into a raised position. Of course, these rollers 182 and 184 swing back into a position to the frame 160 when the piston rod 189 of the double action hydraulic cylinder 187 is retracted. In operation, the installation and removal assembly 2 is rolled out onto the service rails 19 in the undervessel cavity 11 by means of a transition rail 33. At this juncture the cradle 4 is folded into the interior of the carriage 3 in the horizontal position illustrated in FIG. 2B. The maintenance personnel next determine the location of the control rod 7 that needs to be serviced. Once this location is determined, the hydraulic motor of the carriage drive train assembly 48 is actuated in order to turn the carriage wheels 30a, 30b sufficiently to place the assembly 2 at the proper radial position along the service rails 19. The service rails 19 are then rotated along the circumferential rail 31 in order to align the pivot joint 39 adjacent to the control rod drive 7 to be replaced. Next, the hydraulic fluid present within the hydraulic cylinder 43 is slowly released :n order to allow the large bottom length L2 of the carriage 4 to swing down into a vertical position such as that illustrated in FIG. 1. During this step, the hydraulic cylinder 43 advantageously acts as a shock absorber that smoothly and slowly allows the bottom length L2 of the carriage 4 to swing down as the short top length L1 swings upwardly in much the same way as the pneumatic door opener. The lead screw drive assembly 56 of the lifting and lowering mechanism 54 is then actuated to move the retaining bracket 50 all the way up to the distal end of the cradle 4. Next, the kickout cylinder 64 is actuated in order to swing the two stage hydraulic cylinder 58 from a position within the interior 60 of the cradle 4 to a position underneath the retaining bracket 50. Pressurized hydraulic fluid is then admitted to the two stage hydraulic cylinder 58 so that piston rod 139 lifts the retaining cup 59 up through the opening 51 within the bracket 50 and up into a position just under the conical end of a control rod drive 7 to be serviced. The bolts retaining the control rod 7 are then removed. The removal of the last one or two bolts acts to slowly lower the conical end of the control rod drive 7 into the retaining cup 59. The hydraulic cylinder 58 is next slowly relieved of the pressurized hydraulic fluid in its interior, which results in a retraction of the cup 59. As soon as the cup 59 is lowered under the extendible and retractible jaw assembly 52, jaw-clamping cylinder 109 is actuated to open the jaws 105a and 105b, extension and retraction cylinder 122 is actuated to extend the jaws 105a and 105b around the cylindrical body of the control rod drive 7, and jaw clamping cylinder 109 is actuated to pivot the jaws 105a, 105b inwardly and around the control rod drive 7. As the cup 59 passes back through the opening 51 in the retaining bracket 50, the conical end of the control rod drive 7 is captured within the bracket 50 After the piston rod 139 of the cylinder 58 has been allowed to completely retract, the cylinder 58 is pulled back within the interior 60 of the cradle 64 by actuating the kickout cylinder 64 to withdraw piston rod 154. The lead screw drive assembly 57 is then actuated again to move the bottom end of the control rod drive all the way to the bottom end of the cradle 40. Pressurized hydraulic fluid is then readmitted into the hydraulic cylinder 43 of the pivot drive 41, causing the cradle 4 to pivot back into the horizontal position illustrated in FIG. 2B, where upon transfer cart 5 is slid beneath the cradle 4 upon the lower tracks 23 of the service rails 19. Finally, the lifting mechanism 6 of the transfer cart 5 is actuated to draw the rollers 182 and 184 into contact with the control rod drive 7 where upon the jaw clamping cylinder 109 is actuated to release the jaws 105A and 105B from around the control rod drive 7, and the extension and retraction cylinder 122 is actuated to completely withdraw the jaw mechanism 103 from around the body of the drive 7. At the juncture, the motor of the lead screw drive assembly 56 may be actuated momentarily to make sure that the bracket 50 is completely withdrawn from the conical end of the control rod drive 7. The old control rod drive 7 is then wheeled out of the undervessel cavity 11 by the cart 5, where upon a new control rod drive 7 is placed onto the rollers 182 and 184. This new control rod drive is then wheeled beneath the carriage 4, and the process is repeated in reverse to install a new control rod drive 7 within a drive housing 8. |
claims | 1. A method for predicting corrosion rates of a material under a plurality of service conditions, the method comprising:a) determining a first phase composition of the material prior to corrosion;b) exposing the material to a chemical environment representing a first service condition, wherein the environment defines a redox value;c) applying a series of fixed electrical potentials to the exposed material to represent the redox value;d) identifying ranges of the applied potential that correspond to different corrosion behaviors of the material;e) quantifying current and surface electrical properties during corrosion; andf) determining a second phase composition of the material after corrosion in a second service condition. 2. The method as recited in claim 1, wherein the steps of determining the initial phase composition and the second phase composition of the material comprises examining the material with scanning electron microscopy and associated energy-dispersive X-ray emission spectroscopy. 3. The method as recited in claim 1, wherein the step of identifying ranges of the applied potential comprises examining the corroded material with scanning electron microscopy. 4. The method of claim 1 wherein the service conditions are chemical parameters selected from the group consisting of pH, solution concentrations of predetermined moieties, radiolytic products, temperature, and combinations thereof. 5. The method as recited in claim 1 the step of selecting electrical potentials comprises conducting an initial potentiodynamic scan of the material in the chemical environment. 6. The method of claim 5, wherein the applied potential is selected to characterize a material corrosion behavior selected from the group consisting of cathodic corrosion, active corrosion, passive corrosion, or transpassive corrosion. 7. The method as recited in claim 1 wherein the step of quantifying the corrosion comprises applying potentiostatic tests to the material at a plurality of applied potentials. 8. The method as recited in claim 7 wherein electrical impedance spectroscopy is performed during and after the potentiostatic tests to measure the material's electrical properties selected from the group consisting of current, impedance, passivation, and combinations thereof. 9. The method as recited in claim 7 wherein the potentiostatic tests are conducted at several fixed voltages to measure evolution of corrosion current as the material corrodes or stabilizes over time under fixed chemical conditions. |
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description | This application is a continuation of U.S. Non-provisional patent application Ser. No. 13/051,507, filed on Mar. 18, 2011, entitled “Systems and Methods Providing Electron Beam Writing to a Medium,” which is hereby incorporated by reference in its entirety. The present disclosure relates generally to semiconductor manufacturing. Specifically, the present disclosure relates to systems and methods that write to a medium using electron beams. Electron-beam (or “e-beam”) writing relates to a process for creating changes in a medium using e-beams. Specifically, some e-beam processes use e-beams to write designs onto mediums. Examples of mediums that can be written on with e-beams include semiconductor wafers and photomasks (e.g., fused silica and chrome masks). E-beam writing provides a way to create features on a medium where the features are smaller than a resolution limit for light. Some conventional systems use a single-beam method to write designs to a photomask. In one conventional system, in order to mitigate the beam-stitching effect, multiple passes are made by a single beam to apply the desired dosages to the medium. Dosage refers to the amount of electron beam exposure at a given point or area, e.g., e-beam current multiplied by exposure time at a given area is a way to measure dosage. Assuming that the beam is kept at a constant current, dosage increases with a number of passes over an area. Furthermore, throughput is typically inversely proportional to dosage applied by a particular pass. Single-beam exposure methods may be undesirably slow for some applications; thus some applications are evolving to a massive beam exposure technique. Conventional massive beam exposure techniques employ a single source with multiple apertures to generate parallel beams, where each of the parallel beams are individually controllable as to placement, size, dose, and blur. Also, the beams can be individually calibrated. In one conventional technique, a set of parallel beams are used to write parallel strips on a medium simultaneously. The beams are moved in the x-direction by deflection and in the y-direction by scanning movement of the medium to make a zigzag movement to apply a desired dosage and create the parallel strips. However, one issue with conventional massive beam techniques is beam-to-beam variation, and without some way to ameliorate beam-to-beam variation, one or more of the strips may be different from other strips and/or deviate from the desired dosage. Precise calibration for all beams can be difficult, so some conventional techniques account for beam-to-beam variation by overlapping the writing zones between adjacent beams. The overlapped writing zones are referred to as stitches, and while not considered part of the strips, stitches are used to average beam-to-beam variation between adjacent beams. The massive beam techniques can use Gaussian beams, where each beam is a single beam, or patterned beams, where each beam includes a set of sub-beams that are not individually controllable and are arranged in an array. The above-described conventional techniques have some disadvantages. For instance, as mentioned above, techniques using single beams with multiple passes may be undesirably slow, i.e., throughput may not be high enough for some applications. Also, some conventional massive beam techniques using stitching may find throughput negatively affected by the time used to write in the overlapped areas. More efficient and effective e-beam writing is called for. The present disclosure provides for many different embodiments. In a first embodiment, a method for electron-beam writing to a medium includes positioning the medium within an e-beam writing machine so that the medium is supported by a stage and is exposed to an e-beam source. The method also includes writing a pattern to the medium using a plurality of independently-controllable beams of the e-beam source, in which the pattern comprises a plurality of parallel strips. Each of the parallel strips is written using multiple ones of the independently-controllable beams. In another embodiment, an electron-beam writing system includes a stage upon which a medium may be placed and a writing mechanism to write upon the medium placed upon the stage. The writing mechanism includes an electron beam source operable to produce N independently-controllable beams, where N is an integer larger than 1. The system also includes a computer-based control system operable to write a pattern upon the medium in a plurality of parallel strips, each of the strips being written using multiple ones of the N independently-controllable beams. In another embodiment a method for electron-beam writing to a medium includes positioning an e-beam source to write a plurality of strips onto the medium using N independently-controllable beams and writing to each of the strips using multiple ones of the N independently-controllable beams. Variations among the respective independently-controllable beams are averaged by writing to each strip using a unique subset of the N independently-controllable beams. The present disclosure relates generally to semiconductor manufacturing. Specifically, the present disclosure relates to e-beam writing systems and methods that employ multiple beams. While the examples herein discuss applying the techniques to write to photolithographic masks and semiconductor wafers, it is understood that the scope of embodiments can include any system for writing to any appropriate medium using e-beam technology. The following disclosure provides many different embodiments, or examples, for implementing different features of the invention. Specific examples of components and arrangements are described below to simplify the present disclosure. These are, of course, merely examples and are not intended to be limiting. In addition, the present disclosure may repeat reference numerals and/or letters in the various examples. This repetition is for the purpose of simplicity and clarity and does not in itself dictate a relationship between the various embodiments and/or configurations discussed. With reference now to the figures, FIG. 1 is a simplified diagram of a network system 100 in which embodiments may be implemented. Network system 100 includes a network 102 that provides a medium for communications between various devices and computers communicatively coupled by network system 100. Network 102 may be implemented as one or more of various networks, such as the Internet, an intranet, a local area network, a wide area network (WAN), or another network architecture that facilitates communications between network connected devices. Network 102 may include any one or more of various communication connections, such as wire, wireless, fiber optic, satellite links, or other communication media. In the present example, various servers 110-112 are connected to network 102. In addition, a client 120 is connected to network 102. Servers 110-112 may be implemented as hypertext transfer protocol (HTTP) servers, file transfer protocol (FTP) servers, application servers, or other devices that provide data sources such as web pages or other content to client 120 connected therewith. Client 120 may be implemented as a personal computer, a portable computer, a network computer, a super computer, or another computational device. Client 120 is connected to e-beam writing hardware 130, which receives files from client 120 and writes to a physical medium according to the data in the files. The data in the files includes layout patterns that are stored and transmitted in a layout description language such as GDS-II or OASIS. E-beam writing hardware 130 in this example includes writing mechanism 131, which includes a beam source and is operable to produce multiple beams simultaneously, and stage 132, which supports the medium. Either or both of stage 132 and writing mechanism 131 may move during a writing process to facilitate patterning. For purposes of the examples below, stage 132 is moved in the y-direction, and x-direction motion is provided by deflection, though the scope of embodiments is not so limited, as other methods of writing motion are possible. Computers, such as client 120 and/or servers 110-112 may provide control and data to e-beam writing hardware 130 to facilitate writing of designs to the medium. For instance, one or more computers may control the movement of stage 132 and/or deflection during writing processes. FIG. 2 is a simplified block diagram of a computer system 200, such as client 120 or any of servers 110-112 shown in FIG. 1, that may be configured to implement embodiments of an e-beam writing system. Computer system 200 includes a processor 202 interconnected with a system bus 204. System bus 204 provides couplings to subsystems and components of computer system 200. A memory controller 206 interconnected with a system memory 208 provides a communicative coupling between memory 208 and processor 202. Memory 208 may store executable instructions that provide writing functionality as described more fully below. An input/output bridge 210 may be connected with system bus 204, and one or more input/output devices may be connected with an I/O bus 212. For example, a hard disk 216 (or other memory, such as a flash drive) may provide non-transitory, non-volatile storage, and a modem or network adapter 214 may provide a communication interface that facilitates communication exchanges between computer system 200 and one or more data resources on a network. Additionally, user input devices, such as a mouse/keyboard 218, may be coupled with I/O bus 212 and facilitate user input to computer system 200. The configuration of computer system 200 is illustrative and is chosen only to facilitate an understanding of embodiments described herein. FIG. 3 is an illustration of exemplary process 300 for performing e-beam writing according to one embodiment. Process 300 may be performed, for example, by one or more computers 110-112, 120 and e-beam writing hardware 130 of FIG. 1. In block 305, the medium is positioned within the e-beam writing machine so that it is supported by the stage and exposed to the e-beam writing source. The system writes a pattern to the medium, where the pattern is made of multiple parallel strips. Block 305 may further include calibrating the e-beam source so that it can produce N independently-controllable beams, where N is an integer larger than one. Furthermore, block 310 may also include setting a placement, size, and dose for each of the beams. In block 315, writing is performed on the medium using the N independently-controllable beams. In this example, each of the parallel strips is written using more than one of the beams, though perhaps fewer than all N beams. Further in this example, each of the strips is an area of the medium with a single e-beam dosage, the effects of beam-to-beam variation notwithstanding. Example strips are shown in FIGS. 4 and 5. It is understood that beam-to-beam variation may cause dosage to vary somewhat within the bounds of a single strip; however, the use of multiple beams to write to a given strip mitigates the variation within a single strip and among multiple strips. The e-beam writing process includes writing to a physical medium. In one example, the e-beam writing process is used to etch material from a semiconductor wafer in a design that facilitates the manufacture of structures on the wafer. In an other example, the e-beam writing process is used to remove portions of a photomask that is made of, e.g., chrome and fused silica. The photomask can then be used in the manufacturing process of semiconductor devices. E-beam writing processes provide an effective way to create designs on a physical medium where some of the features of the design may be too small to be made by other conventional processes, such as photolithography. The scope of embodiments is not limited to the example shown in FIG. 3. Other embodiments may add, omit, rearrange, or modify actions. For instance, some embodiments may repeat the actions of FIG. 3 many times to write multiple complex structures to the medium. Furthermore, other embodiments include subsequent processing steps appropriate for a photomask, semiconductor wafer, or other medium. As shown below, some example embodiments include use of deflection to write a zigzag path using each of the beams. In other embodiments, either scanning or deflection is used to provide relative motion for writing. Still further, some embodiments employ Gaussian beams, whereas other embodiments employ patterned beams. Moving to FIG. 4, a medium 400 with strips 401 thereon is depicted according to one embodiment during e-beam writing. Dot 402 represents a placement at a point in time of one e-beam, and other dots in FIG. 4 represent other e-beams similarly. In FIG. 4, the area of each strip 401 is double exposed. Lines 410 show an example path traced by the e-beam of strip 401a. In FIG. 4, the stage (not shown) moves in the y-direction, and deflection is used to create relative movement in the x-direction. The result is the path represented by lines 410. The dot adjacent dot 402 represents a beam that also writes to strip 401a in a manner similar to that described immediately above. The two beams together write strip 401a. Portions 404 and 405 are shown as not double exposed, though various embodiments include techniques to ensure that portions 404, 405 are double exposed. For instance, dummy exposures may be used to double expose portions 404, 405. Also, the stage can be used to position the medium to make further exposures where desired to provide consistent double exposure throughout medium 400. In some embodiments, the technique illustrated in FIG. 4 brings superior beam-to-beam uniformity than conventional stitching techniques discussed above. Of note in the embodiment of FIG. 4 is that each strip 401 is created using a unique set of beams. In some embodiments, throughput is not adversely affected by tying up double the beams on a single set of strips. If the desired dose for a strip 401 is one unit, the beams can be kept at one current unit while doubling writing speed to two speed units. Since each strip 401 is double exposed, the cumulative dose for each strip is one unit, and it is performed in one-half time unit because the writing velocity is doubled. The beams can then be moved to a different but similarly-sized portion of the medium to write a unit dose in another one-half time unit. Thus, the beams can be fully utilized in a given time unit. A rule of thumb for some embodiments is that velocity is increased by a factor equal to the exposure factor so as to achieve the same throughput. FIG. 5 is an illustration of portion 500 of a medium according to one embodiment during e-beam writing. In the example of FIG. 5, strips 501 are quadruple exposed by the end of the process, and line 510 shows an example path written by a beam that is represented by dot 502. Strips 501a, 501c, and 501d are shown as having less than quadruple exposure. Only strip 501b falls within quadruple exposure area 501, whereas strips 501a, 501c, and 501d fall within exposure areas 504, 505 with less than quadruple exposure. As explained above with respect to FIG. 4, various techniques may be used to provide consistent quadruple exposure. For instance, in one example, dummy exposures may be added to strips 501a, 501c, and 501d. Additionally and/or alternatively, the beams can be moved to the right or to the left adjacent the end beam on that side, where writing is begun. For instance, in the present example, there are ten beams represented by ten dots 502. After the beams have traversed the medium, the stage can be moved to position the medium so that the ten beams are moved in the x-direction ten places and begin writing to the immediate right (or left) of the original ten beams, thereby writing additional strips and providing quadruple exposure to areas 504 (and/or 505). Such operation can be performed in any embodiment using any number of beams per strip. Once again, the operation illustrated by FIG. 5 may be performed to have no negative impact on throughput. For instance, similarly to the operation described above for FIG. 4, the writing speed can be increased by the exposure factor. In fact, in the embodiments of FIGS. 4 and 5, the writing speed can be changed by any arbitrary factor to increase or decrease throughput as appropriate. FIG. 6 is an illustration of exemplary beam arrangement 600 for use with an embodiment. Beam arrangement 600 includes four independently-controllable beams 610, 620, 630, 640. Each of the independently-controllable beams 610, 620, 630, 640 includes an M×M array of sub-beams, where M is an integer greater than one (and in this case is equal to seven). In other embodiments, M can be any arbitrary integer greater than one. The sub-beams are illustrated in FIG. 6 by exemplary sub-beams 611 (in beam 610), 621 (in beam 620), 631 (in beam 630), and 641 (in beam 640). The sub-beams themselves are not independently-controllable. Beam 640 is overlaid by beam 630, which is overlaid by beam 620, which is overlaid by beam 610. Between each adjacent beam there is an x-y offset so that the beams 610, 620, 630, 640 do not lay directly on top of each other. Furthermore, there is an angular offset, alpha, with respect to a direction of scanning (in this case, it assumed that the direction of scanning is along the y-direction). Furthermore, in this embodiment, x-direction movement is not used when scanning a given strip, so that the movement relevant to the discussion of FIGS. 6 and 7 is the y-direction movement. FIG. 7 is an illustration of beam arrangement 600 shown to emphasize the individual pixels covered by the sub-beams, where each dot in FIG. 7 represents a sub-beam. The dots in FIG. 7 are an aggregation of the four 7×7 arrays of sub-beams shown in FIG. 6, for a total of 196 sub-beams. FIG. 7 also illustrates an exemplary scan direction and an angular offset. During writing, the beams are projected onto the medium as shown in FIGS. 6 and 7, and the medium is moved relative to the beam sources to provide scanning movement. The beams 610, 620, 630, 640, are performed simultaneously in this example. View 710 is an illustration of the projection of the pixels along a line perpendicular to the scan direction. The x-y offset and angular offset produce a configuration wherein adjacent pixels are from different beams. In the case of view 710, the first pixel is from beam 610, the second pixel is from beam 620, the third pixel is from beam 630, and the fourth pixel is from beam 640. The placement then repeats across the view in the x-direction. View 710 illustrates that a strip written by beam arrangement 600 includes contributions from four different beams 610, 620, 630, 640, and the sub-beams of those beams are distributed in a way that averages out the beam-to-beam variation among beams 610, 620, 630, 640. FIG. 8 is an exemplary pixel projection along the x-direction according to one embodiment consistent with the examples above for FIGS. 6 and 7. In the embodiments of FIGS. 6-8, each strip is covered by multiple independent beams, and adjacent pixels are exposed by different beams. In FIG. 8, strip 1 is exposed by a set of four independent beams, and strip 2 is exposed by a different set of four independent beams. While not shown in FIG. 8, it is noted that different sets of beams can be used to expose either or both of strips 1 and 2 in subsequent exposures. The subsequent exposures can further reduce the effects of beam-to-beam variation and also enable gray-level writing. In FIG. 8, the x-y and angular offsets combine to make a cumulative offset substantially equal to one pixel width. The one-pixel-width offset creates the pattern shown in FIGS. 7 and 8, where the pixels have a repeating pattern along the x-direction (perpendicular to the y-direction scanning). Other embodiments may use a cumulative beam-to-beam offset that is greater than one pixel width. FIG. 9 is an illustration of an exemplary x-direction projection according to one embodiment that uses a beam-to-beam offset greater than one pixel width. The combination of beams to cover a given strip is not fixed or regular in the example of FIG. 9. Each strip is exposed by different combinations of individual beams, which further mitigates the effects of beam-to-beam variation. Strips 1 and 3 are shown as being not exposed with the same dose as strip 2, but it is understood that the number of strips extends beyond strip 1 and strip 3, and beams writing those additional strips (not shown) add pixels to strips 1 and 3 to apply a consistent exposure. The examples above in FIGS. 6-9 show example embodiments using four or eight beams, though the scope of embodiments is not so limited. Rather, various embodiments may include any arbitrary number of beams. Furthermore, x- and y-directions are used purely for illustration and do not limit the various embodiments to any particular orientation. FIG. 10 is an illustration of a 7×7 array 1000 of sub-beams for use with the embodiments of FIGS. 6-9. Pb is a spacing between sub-beams in array 1000. Pproj is a width of a sub-beam as it is projected onto the medium. Alpha is the angular offset. In this example, a pixel size is 3.5 nm. The equations below show the area covered per scan in this example. It is noted that the e-beam current and the writing speed can be adjusted in view of the below equations to have a desired effect on throughput. The numbers below are for illustration only and do not limit the scope of embodiments. consider P proj is the multiple of 3.5 nm pixel size P proj = P b · sin ( α array ) = 10.5 nm = 3 × 3.5 nm ⇒ P b = 74.25 nm Beam Shift = 1 pixel If w proj = Strip Width N subbeans along w proj = w proj PixSize = 504 nm 3.5 nm = 144 N beams for 1 Strip = N subbeams along w proj N subbeams array = 144 7 = 21 N strip per Scan = Total Beam Number N beams for 1 Strip = 13000 21 = 619 Area Covered per Scan = N strip per Scan × w proj = 619 × 504 nm = 0.312 mm Various embodiments may include advantages over other techniques. For instance, various embodiment write to each strip using multiple beams, thereby helping to average out beam-to-beam variation without stitching. It is understood that the embodiments illustrated above may not in all scenarios completely eliminate beam-to-beam variation. However, it is enough for many applications that the beam-to-beam variation is mitigated by use of multiple beams per strip. Furthermore, as explained above, parameters (e.g., e-beam current and writing speed) can be adjusted so that the effect on throughput is not negative in some instances. Additionally, some embodiments perform gray-level writing by applying the multi-pass writing technique described above. The foregoing has outlined features of several embodiments so that those skilled in the art may better understand the detailed description that follows. Those skilled in the art should appreciate that they may readily use the present disclosure as a basis for designing or modifying other processes and structures for carrying out the same purposes and/or achieving the same advantages of the embodiments introduced herein. Those skilled in the art should also realize that such equivalent constructions do not depart from the spirit and scope of the present disclosure, and that they may make various changes, substitutions and alterations herein without departing from the spirit and scope of the present disclosure. |
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claims | 1. A precision robot arm system comprising:a two-dimensional (2D) vertical plane robot arm including duo-vertical-stages and a kinematic linkage system supporting an X-ray detector for positioning and manipulating of the X-ray detector and including;a first cross-roller bearing structure and a second cross-roller bearing structure carried by said 2D vertical plane robot arm; anda vertical support carrying said 2D vertical plane robot arm, said vertical support including spaced apart rails respectively engaging said first cross-roller bearing structure and said second cross-roller bearing structure carried by the 2D vertical plane robot arm, such that said 2D vertical plane robot arm is rotatable and slidingly moved along said vertical support. 2. The precision robot arm system as recited in claim 1 wherein said X-ray detector supported by said 2D vertical plane robot arm includes a 2D X-ray pixel detector. 3. The precision robot arm system as recited in claim 1 includes a granite base with an air-bearing support supporting said vertical support carrying said 2D vertical plane robot arm. 4. The precision robot arm system as recited in claim 1 includes a 2D horizontal base stage mounted on a granite base with an air-bearing support supporting said vertical support carrying said 2D vertical plane robot arm. 5. The precision robot arm system as recited in claim 1 includes a vertical axis goniometer supporting said vertical support carrying said 2D vertical plane robot arm. 6. The precision robot arm system as recited in claim 1 wherein said 2D vertical plane robot arm includes a 3D fast scanning stages group provided with said 2D vertical plane robot arm, said 3D fast scanning stages group performs detector raster scans. 7. The precision robot arm system as recited in claim 1 wherein said 2D vertical plane robot arm provides effective positioning and manipulating of said X-ray detector at an X-ray nanoprobe beamline. 8. The precision robot arm system as recited in claim 1 said 2D vertical plane robot arm provides a set load capacity with micron level positioning repeatability. 9. A precision robot arm system comprising:a two-dimensional (2D) vertical plane robot arm including duo-vertical-stages and a kinematic linkage system;said 2D vertical plane robot arm including a first high rigidity cross-roller bearing structure and a second high rigidity cross-roller bearing structure; anda robot arm vertical support carrying said 2D vertical plane robot arm; said vertical support including spaced apart rails respectively engaging said first high rigidity cross-roller bearing structure and said second high rigidity cross-roller bearing structure carried by the 2D vertical plane robot arm, such that said 2D vertical plane robot arm is rotatable and slidingly moved along said vertical support. 10. The precision robot arm system as recited in claim 9 further comprising a 2D horizontal base stage supporting said robot arm vertical support carrying said 2D vertical plane robot arm, said 2D horizontal base stage for precisely positioning said 2D vertical plane robot arm. 11. The precision robot arm system as recited in claim 9 further comprising a 3D fast scanning stages group provided with said 2D vertical plane robot arm for moving and precisely positioning said 2D vertical plane robot arm, said 3D fast scanning stages group performs detector raster scans. 12. The precision robot arm system as recited in claim 9 wherein said first bearing structure and said second bearing structure include a respective bearing formed of a stainless steel material. 13. The precision robot arm system as recited in claim 9 wherein said 2D vertical plane robot arm is formed of a selected material of aluminum, natural granite, synthetic granite, carbon fiber, and a combination of thereof. 14. The precision robot arm system as recited in claim 9 wherein said 2D vertical plane robot arm carries an X-ray detector providing effective positioning and manipulating of said X-ray detector at an X-ray nanoprobe beamline. 15. A method for implementing precision robot arm system comprising:providing a two-dimensional (2D) vertical plane robot arm including duo-vertical-stages and a kinematic linkage system;providing the 2D vertical plane robot arm with a first high rigidity cross-roller bearing structure and a second high rigidity cross-roller bearing structure; andproviding a vertical support member carrying the 2D vertical plane robot arm; said vertical support including spaced apart rails respectively engaging said first high rigidity cross-roller bearing structure and said second high rigidity cross-roller bearing structure carried by the 2D vertical plane robot arm, such that said 2D vertical plane robot arm is rotatable and slidingly moved along said vertical support. 16. The method as recited in claim 15 wherein providing the 2D vertical plane robot arm includes providing a 2D horizontal base stage supporting said robot arm vertical support carrying said 2D vertical plane robot arm, said 2D horizontal base stage for precisely positioning said 2D vertical plane robot arm. 17. The method as recited in claim 15 wherein providing the 2D vertical plane robot arm includes providing a 3D fast scanning stages group with said 2D vertical plane robot arm for moving and precisely positioning said 2D vertical plane robot arm, said 3D fast scanning stages group performs detector raster scans. |
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claims | 1. Apparatus for placing and moving a spacer (300) into and along an annulus (150) defined by inner and outer coaxial cylindrical tubes (100, 200), said apparatus comprising:a first push rod (1000);a radially expandable generally toroidal pusher (600) surrounding and located at a first axial position on said push rod (1000);a collet and mandrel (700) operatively associated with the push rod (1000) axially adjacent the pusher (600) for engaging an end of said inner cylindrical tube (100) and providing a generally frusto-conical surface from the outer diameter of the push rod (1000) to the outer diameter of the inner cylindrical tube (100);a second push rod (1006) supported for axial movement by said push rod (100);a magnetic ram operatively associated with the second push rod (1006) for movement therewith from a first position located radially inwardly of the pusher (600) to a second position located within the inner cylindrical tube (100) for creating a magnetic coupling with the pusher (600) and moving the pusher from said first location on the push rod (1000) into the annulus (150); andwherein the pusher (600) comprises a plurality of arcuate ferromagnetic segments (601, 603) forming a generally toroidal, radially expandable ferromagnetic ring surrounding the push rod (1000) and normally located axially adjacent the fixed end (708) of the collet and mandrel (700) movable under the influence of a magnetic field from the electromagnetic coil (500) along the collet and mandrel (700) toward a radial shoulder (706) on a collet and mandrel finger (702) and into the annulus (150) between the inner and outer cylindrical tubes (100, 200) and thereafter movable under the influence of a magnetic field from the electromagnetic coil (500) to position a toroidal spacer (600) in said annular space. 2. Apparatus as defined in claim 1 further comprising:a source of electrical current (1200);an electric actuator and drive (1050) electrically connected to the source of electric current (1200);said magnetic ram (400) including a coil (500) electrically connected to the source of current (1200) for generating a magnetic field; andcontrols for switching and regulating the flow of electric current (1200) associated with the source of electrical current (1200) and electrically connected to the electric actuator and drive (1050) and the coil (500). 3. Apparatus as defined in claim 2 further comprising:an axially movable, radially expandable collet (350) mounted on the push rod (1000) and normally located radially inwardly of the collet and mandrel (700) for expanding the collet and mandrel (700) into engagement with the inner cylindrical tube (100). |
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052395642 | abstract | The irradiated component (1) comprises at least one tubular wall with its axis vertical and fastened inside a well (2) in a concrete structure (3). The connecting elements between the concrete structure (3) and the component (1) are destroyed, the component (1) is displaced in successive steps in the vertical direction along its axis on the inside of the well (2), and the wall of the component is cut, so as to obtain blocks (26) of irradiated material of the wall, at the upper level of the well (2) of the concrete structure (3) after each displacement of the component (1). The cut blocks (26) are disposed of for the purpose of elimination or storage. The displacement of the component can be obtained by a pull or push on its lower bottom. |
abstract | An exposure apparatus includes a chuck for holding an object and an optical system for directing light from a light source to the object held by the chuck. The optical system includes a multilayer film mirror which has a concave reflecting surface, wherein an area of contacting portions of the chuck is set so that the contact portions area of the chuck is at most 10% of an area of the object held by the chuck. |
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054935903 | summary | The present invention relates to a critical power enhancement system for a pressurized fuel channel type nuclear reactor by inducing turbulence at strategic locations within fuel bundles of the reactor and thereby improving critical heat flux of the fuel bundles. BACKGROUND OF THE INVENTION An example of a pressurized fuel channel type nuclear reactor is the CANDU.TM. reactor which contains a plurality of pressure tubes defining fuel channels therethrough. Each fuel channel is horizontally oriented and contains a plurality of fuel bundles, generally arranged end to end. Each fuel bundle contains a set of solid fuel rods or elements containing fissionable material. High pressure heavy water coolant enters the fuel channel at one end, flows over the fuel bundles and through the gaps between the fuel elements so as to cool the fuel elements and remove the heat from the fission process, and exits from the fuel channel at the other end. This heat is subsequently transferred by the coolant to a heat exchanger which produces steam that drives a turbine to produce electrical energy. The heavy water flowing in the water gaps is pressurized and does not boil significantly. The maximum power that can be produced within a fuel channel is determined by the maximum power that can be produced safely by individual fuel bundles within that channel. This maximum power within the fuel channel is normally referred to as the Critical Channel Power or CCP. The maximum power that can be produced safely by any given fuel bundle within that channel is called the Critical Bundle Power, and is determined by the variation in power production within that bundle, the corresponding local coolant conditions, and the design of the fuel bundle. The Critical Bundle Power is the power corresponding to the onset of a significant decrease in the efficiency of heat transfer from the bundle to the coolant, and the local heat flux at which this happens is referred to as the Critical Heat Flux or CHF. Since the high temperatures that can occur when the CHF is exceeded may damage the fuel bundle, the channel power and flow conditions are set to ensure that CHF is never exceeded in any bundle. CHF occurs on a heated fuel element when some pan of its surface can no longer be continuously wetted by the liquid coolant. There are two types of CHF, i.e., the departure from nucleate boiling (DNB) type and the liquid depletion type. The actual mechanism for this depends upon the location of the fuel element, and the thermal hydraulic conditions of the coolant surrounding it. In order to ensure the CHF is never exceeded in any bundle, a safety factor or operating margin is applied to the CCP, which in turn results in the power that can be produced by the pressurized fuel channel type of nuclear reactor being reduced by approximately the same factor. If, however, the CHF could be increased, the power that could be produced by the reactor could also be increased. A similar situation also applies to other types of water-cooled nuclear reactors. Improvements of CHF on fuel bundles for pressure vessel type light water reactors have been suggested through the following separate methods: (i) addition of extra grid spacers and mixing vanes (U.S. Pat. No. 4,698,204 issued to Taleyarkhan on Oct. 6, 1987); (ii) installation of a tubular flow diverting channel, flow diverting panels or other flow diverting/deflecting devices (U.S. Pat. No. 4,738,819 issued to Taleyarkhan on Apr. 19, 1988, U.S. Pat. No. 4,678,631 issued to Taleyarkhan on Jul. 7, 1987, U.S. Pat. No. 3,663,367 issued to Calvin on May 16, 1972, and Canadian Patent 1,115,863 issued to the Babcock & Wilcox Company on Jan. 5, 1982); (iii) cavitating the element sheath surface (U.S. Pat. No. 4,474,231 issued to Staub et al. on Oct. 2, 1984); and (iv) installation of a special flow distributing channel within the fuel element assembly (U.S. Pat. No. 4,708,846 issued to Patterson et al. on Nov. 24, 1987). The CHF enhancement methods proposed for pressure vessel type light water reactors are not directly applicable to fuel bundles of pressurized fuel channel type nuclear reactors, such as the CANDU.TM. reactor. This is mainly because of the physical differences between fuel assemblies in the pressure vessel type reactors and the fuel bundles of the pressurized fuel channel type reactors. The pressure vessel type reactor uses very long fuel assemblies which stretch the full length of the reactor, and uses much larger spacing between the individual elements. The type of flow turbulence promoter necessary to provide optimal CHF enhancement in the pressure vessel type reactors, such as extra grid spacers and flow diverting devices, may impose unnecessarily high hydraulic resistance in the fuel channels of the pressurized fuel channel type reactor. Their introduction would also require mechanical changes to the fuel bundle of the pressurized fuel channel type reactor which would affect its overall behaviour. They are therefore not practical for pressurized fuel channel type reactors. Canadian Patent No. 1,115,863 discloses provision of ridges on guide tubes for control rods. This patent specifies that the subject invention is suitable for preventing early occurrence of the departure from nucleate boiling (DNB) type of CHF, which is likely to occur near the gap between adjacent fuel elements, or between guide tubes and fuel elements. The guide tube in Canadian Patent 1,115,863, is effectively unheated, and therefore has a relatively colder surface than the surrounding fuel elements. The ridges are used to strip liquid off the cold guide tube so that it will be available to the hot fuel elements. The guide tubes are a unique feature of fuel assemblies for pressure-vessel-type, water cooled reactors. In addition, liquid depletion type of CHF is more likely to occur than the DNB type of CHF in a pressurized fuel channel type reactors. The invention described in Canadian Patent 1,115,863 is only applicable to the pressure vessel type of reactor and is not applicable to pressurized fuel channel type reactors where there are no guide tubes. A method has been proposed, as shown in FIG. 1, to improve CHF in pressurized fuel channel type reactors by the installation of roughness rings 10 on the inner surface of fuel channels 12 (U.S. Pat. No. 3,372,093 issued to Wikhammer et al. on Mar. 5, 1968). The CHF enhancement is accomplished since the roughness rings 10 create turbulence and redistribute liquid from the unheated channel wall to the fuel bundle 14 supported by grid spacers 16 within the fuel channel 12. This CHF enhancement method proposed for pressurized, fuel-channel type reactors has the following drawbacks: The presence of roughness ring on the inner surface of the fuel channel creates practical difficulties for the passing of fuel bundle within the fuel channel. The provision of the roughness ring on the inner surface of the fuel channel would therefore require mechanical changes to the fuelling system of the pressurized fuel channel type reactors, which is undesirable. In addition, channel roughness rings increases the hydraulic resistance in the fuel channel significantly. As discussed below, an increase in hydraulic resistance in the fuel channel causes the coolant flow to decrease, and hence causes the CHF to occur at a lower fuel channel power. The resulting CCP is either worse than the case without the CHF enhancement devices, or is only improved to an insignificant degree. High hydraulic resistance may also reduce the coolant flow through the fuel channels in an existing reactor that was not designed to accommodate a large pressure-drop resulting from such a large hydraulic resistance, thus affecting the overall performance of the reactor. SUMMARY OF THE INVENTION The present invention seeks to provide a critical power enhancement system for a pressurized fuel channel type nuclear reactor, which improves the critical heat flux of the fuel bundles without significantly increasing the hydraulic resistance in the fuel channel. In accordance with one aspect of the present invention, there is provided a fuel element assembly for use in a nuclear reactor using a pressurized water coolant flowing in a plurality of fuel channels, the fuel element comprising: an elongated fuel element containing fissionable material; a plurality of spacers attached to the fuel element; a plurality of bearing pads attached to certain type of fuel element; and at least one CHF enhancement appendage, attached to the fuel element along its length thereof and projecting outwardly from the surface of the fuel element for generating turbulence in the coolant flowing along the length of the fuel element downstream of the location of the CHF enhancement appendages. The CHF enhancement appendages are functionally different from spacers for separating the fuel elements from each other or bearing pads for providing load bearing to the fuel bundle. In accordance with another aspect of the present invention, there is provided a fuel bundle assembly for use in a nuclear reactor using a pressurized water coolant flowing in a plurality of fuel channels, the fuel bundle assembly comprising: a fuel bundle containing a plurality of fuel elements, each of the fuel elements having a length and ends; and at least one CHF enhancement appendage attached to each of certain fuel elements along the length thereof and projecting outwardly from the surface of the each fuel element for generating turbulence in the coolant flowing along the length of the fuel bundle downstream of the locations of the CHF enhancement appendages. The fuel elements are arranged in a bundle form with both of the ends of the fuel elements mechanically hold together, and separated from each other by a pair of spacers each attached to one of two neighbouring fuel elements. The fuel elements located on the outer ring of the fuel bundle are further provided with bearing pads for providing load bearing to the fuel bundle. In accordance with another aspect of the present invention, there is provided a pressurized fuel channel type nuclear reactor which uses pressurized water as a coolant, the reactor comprising: a plurality of pressure tubes each defining a fuel channel therethrough, the pressure tubes having a length, an inlet for introducing the coolant into the fuel channels, and an outlet for discharging the coolant from the fuel channels; a plurality of fuel bundles contained within each of the pressure tubes, each of the fuel bundles containing a plurality of fuel elements therein, each of the fuel elements having a length and ends, the length of each fuel element being parallel to the length of the pressure tube; and at least one CHF enhancement appendage is provided on certain fuel elements along the length thereof and projecting outwardly from the surface of the fuel element for generating turbulence in the coolant flowing along the length of the fuel bundle downstream of the locations of the CHF enhancement appendages. The fuel elements are arranged in a bundle form with both of the ends of the fuel elements mechanically held together and separated from each other by a pair of spacers each attached to one of two neighbouring fuel elements. The fuel elements located on the outer ring of the fuel bundle are further provided with bearing pads for providing load bearing to the fuel bundle. In accordance with another aspect of the present invention, there is provided a critical power enhancement system for a pressurized fuel channel type nuclear reactor which uses pressurized water as a coolant, the reactor comprising: a plurality of pressure tubes each defining a fuel channel therethrough, the pressure tubes having a length, an inlet for introducing the coolant into the fuel channels, and an outlet for discharging the coolant from the fuel channels; a plurality of fuel bundles contained within each of the pressure tubes, each of the fuel bundles containing a plurality of fuel elements therein, each of the fuel elements having a length and ends, the length of each fuel element being parallel to the length of the pressure tube; the critical power enhancement system comprising a plurality of CHF enhancement appendages attached to certain fuel elements along the length thereof and projecting outwardly from the surface of the fuel element into the coolant flow space within the fuel channel surrounded by fuel elements, for generating turbulence in the coolant flowing along the length of the fuel bundle downstream of the locations of the CHF enhancement appendages. The fuel elements are arranged in a bundle form with both of the ends of the fuel elements mechanically hold together and separated from each other by a pair of spacers each attached to one of two neighbouring fuel elements. The fuel elements located on the outer ring of the fuel bundle are further provided with bearing pads for providing load bearing to the fuel bundle. The locations of the CHF enhancement appendages within the fuel bundle are such that the turbulence generated by the CHF enhancement appendages occurs at locations within the fuel bundle where CHF is most likely to occur. |
claims | 1. A tube support structure for use within an array of heat exchanger tubes, each tube having a diameter and a longitudinal axis, the tube support structure comprising:a tube support bar for use between a pair of heat exchanger tubes, the support bar comprised of a first metallic strip attached to a second metallic strip at spaced intervals, the first strip having a coefficient of thermal expansion greater than the second strip, wherein the first strip is attached to the second strip at spaced intervals in a direction transverse to the length of the second strip, and wherein the first strip is attached to the second strip at spaced intervals of about 2 tube diameters. 2. The tube support structure of claim 1, wherein the first and second strips are flat at a first temperature, and wherein the first strip becomes convex at a temperature higher than the first temperature. 3. The tube support structure of claim 1, wherein the first strip is thinner than the second strip. 4. The tube support structure of claim 3, wherein the first strip is 0.02 inches thick and the second strip is 0.08 inches thick. 5. The tube support structure of claim 1, wherein the first strip is attached to the second strip via spot welding. 6. The tube support structure of claim 1, wherein the tube diameter is about 0.5 inches and the first strip is attached to the second strip at spaced intervals of about 1 inch. 7. The tube support structure of claim 1, wherein the first strip is made of SB-166 1690 and the second strip is made of SA 240 type 410S. 8. The tube support structure of claim 1, wherein said tube support structure is operational at a temperature of about 550 degrees F. 9. The tube support structure of claim 1, wherein the tube support bar is a low-bar of a lattice tube support bar array. 10. The tube support structure of claim 1, wherein the tube support bar is a high-bar of a lattice tube support bar array. 11. A tube support structure for use within an array of heat exchanger tubes, each tube having a diameter and a longitudinal axis, the tube support structure comprising:a tube support bar for use between a pair of heat exchanger tubes, the support bar comprised of a first metallic strip attached to a second metallic strip at spaced intervals, the first strip having a coefficient of thermal expansion greater then the second strip,a third metallic strip attached to the second metallic strip opposite the first metallic strip at spaced intervals, the third strip having a coefficient of thermal expansion greater than the second strip. 12. A support for heat transfer tubes in a steam generator, the support comprising:a plurality of bars installed between the heat transfer tubes so that a gap exists between the bars and the heat transfer tubes;a spring means welded to at least one of the bars at intervals, the spring means having a thinner thickness than the bar; andwherein the spring means and the bar have different thermal expansion coefficients so that at a non-operating temperature of the steam generator the spring means does not contact the adjacent tube and at the operating temperature of the steam generator the spring means contacts the adjacent heat transfer tube. |
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summary | ||
047605901 | abstract | A multioperation accelerator simulating more especially cobalt radiation. According to the invention, an electron accelerator of average energy and a set of targets and filters are provided so as to obtain at least one penetration curve similar to that of cobalt and other closely related curves. |
abstract | Non-intrusive error detection techniques for control and shutdown rod position in nuclear reactors, including methods of monitoring digital rod position indication (DRPI) signals of a DRPI system of a nuclear power plant. The methods include acquiring digital rod position signals at a point between a DRPI display cabinet and a DRPI data cabinet of the DRPI system, and processing the digital rod position signals to identify variations in a signal level and a signal timing of the digital rod position signals to determine rod position errors of the DRPI system. |
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048809896 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to shielding apparatus for radioactive materials. In particular, the invention relates to shielding apparatus utilized with radioaerosol delivery systems in nuclear medicine. 2. State of the Art Lung ventillation scanning using radiolabeled aerosols has been studied for about the last 20 years. However, until recently when improved aerosol generating devices have become more generally available, practical applications of such methods have been extremely limited. One particularly useful aerosol generating system is that described in U.S. Pat. No. 4,116,387 and U.S. Pat. No. 4,251,033, the disclosures of which are incorporated herein by reference. The nebulizer described in those patents has been found to be particularly useful in generating aerosols having a particle size and particle size distribution to make lung scanning a useful diagnostic tool. Relatively recent articles describing lung scanning methodology utilizing radioactive aerosols can be found at: Radiology, 131:256-258, April 1979; Seminars in Nuclear Medicine, Volume X, No. 3 (July), 1980, pp. 243-251; and The Journal of Nuclear Biology and Medicine, Vol. 19, No. 2, 1975, pp. 112-120. Because of the increased interest in using radioaerosols for diagnostic imaging, there is a need for a compact and practical apparatus for delivering such radioaerosols to a patient. The invention described herein is considered to satisfy such a need. SUMMARY OF THE INVENTION The present invention is directed in one aspect to an apparatus comprising support means for supporting a radioaerosol generating source; transport means connectable to the source for transporting a radioactive aerosol generated by the source to a patient in fluid communication with the source; and shielding means substantially surrounding the support means and the transport means for reducing the amount of radiation transmitted to the surroundings, a portion of the shielding means being releasably attachable to the transport means and being removable with said transport means and said source from the support means as a unit. In another aspect, the invention is directed to a shielding container comprising an outer shell; an inner shell supported within the outer shell, the inner shell being formed with an inner wall and an outer wall defining a space therebetween for receiving radiation shielding material, the inner wall having a portion thereof conforming substantially to the contours of a radioaerosol source and transport means to be placed therein; and a removable cover formed with radiation shielding material and having a portion thereof conforming generally to the contours of the radioaerosol transport means to be positioned thereunder, the inner wall and the cover defining at least one opening therebetween to permit the transport means placed therein to be in fluid communication with the surrounding atmosphere and/or patient when in use. |
054830648 | claims | 1. A scanning microscope for scanning a sample having a surface disposed substantially perpendicular to a tip of a probe to generate a signal indicative of a property of the sample as a function of planar coordinates along the surface of the sample, comprising: (a) a rigid support structure (b) scanning means having a fixed end mounted on said support structure and a free end movable along a plane substantially perpendicular to a main axis of the scanning means in response to control signals applied to the scanning means; (c) a probe having a tip fixedly disposed in coaxial alignment with said main axis of the scanning means, said tip with a point on the surface of the sample precisely aligned with said tip to produce a signal indicative of said property of the sample at said point, thereby providing a map of such property as a function of position on the surface of the sample; (d) a stand-off slidably mounted on said free end of the scanning means, said stand-off comprising a bottom surface slidably connected to the free end of the scanning means and a top surface comprising means for receiving said sample fixedly mounted thereon; and (e) sample-positioner means mounted on said support structure and comprising means for slidably moving said stand-off over said free end of the scanning means in order to position the sample to a desired location for scanning, said sample-positioner means also comprising means for disengaging the stand-off after positioning of the sample and providing a sufficient gap between the stand-off and the sample-positioner means to prevent contact and resulting interference therebetween during scanning. wherein said scanning means consists of a piezoelectric unit; wherein said sample-positioner means comprises a plate having an opening sufficiently large to surround said stand-off, said plate being disposed coplanarly therewith, the plate comprising means for slidably moving said stand-off in order to position the sample to a desired location for scanning and further comprising means for disengaging the stand-off after positioning of the sample, so as to provide a sufficient gap between the stand-off and the sample-positioner means to prevent contact and resulting interference therebetween during scanning; and wherein said scanning microscope further comprises: apparatus for manually controlling the motion of the sample positioner means in response to a visual signal indicative of the relative position of a desired target point on the surface of the sample with respect to the tip of the probe; and apparatus for automatically controlling the motion of the sample-positioner means to provide said gap between the stand-off and the sample-positioner means in response to a signal indicative of said desired target point on the surface of the sample having been placed in alignment with the tip of said probe. (a) providing a rigid support structure; (b) providing scanning means having a fixed end mounted on said support structure and a free end movable along a plane substantially perpendicular to main axis of said scanning means in response to control signals applied to the scanning means, said free end comprising means for slidably mounting said sample thereon; (c) providing a probe having a tip fixedly disposed in coaxial alignment with said main axis of the scanning means, said tip interacting with a point, the surface of the sample precisely aligned with said tip to produce a signal indicative of a property of the sample at said point; (d) mounting the sample on said free end of the scanning means; (e) slidably moving said sample over said free end of the scanning means in order to position the sample such that a desired target thereon is aligned with said tip of the probe for scanning; and (f) scanning the sample maintaining said tip and said scanning means in substantial coaxial alignment; whereby a map of such property as a function of position on the surface of the sample is provided; (g) providing a stand-off slidably mounted on said free end of the scanning means, said stand-off comprising a bottom surface slidably connected to the free end of the scanning means and a top surface to receive said sample fixedly mounted thereon; and (h) providing sample-positioner means mounted on said support structure and to slidably move said stand-off over said free end of the scanning means in order to position the sample to a desired location for scanning, said sample-positioner means also to disengage the stand-off after positioning of the sample and provide a sufficient gap between the stand-off and the sample-positioner means to prevent contact and resulting interference therebetween during scanning; and wherein step (d) is accomplished by fixedly mounting the sample on said top surface of the stand-off; and step (e) is accomplished by slidably moving said stand-off with said sample-positioner means to position the sample to said desired location for scanning and by thereafter disengaging the stand-off and providing a sufficient gap between the stand-off and the sample-positioner means to prevent contact and resulting interference therebetween during scanning. 2. The scanning microscope recited in claim 1, wherein said scanning means consists of a piezoelectric unit. 3. The scanning microscope recited in claim 2, wherein said piezoelectric unit consists of a piezoelectric ceramic tube. 4. The scanning microscope recited in claim 1, wherein said sample-positioner means comprises a plate having an opening sufficiently large to surround said stand-off, said plate being disposed coplanarly therewith, the plate comprising means for slidably moving said stand-off in order to position the sample to a desired location for scanning and further comprising means for disengaging the stand-off after positioning of the sample, so as to provide a sufficient gap between the stand-off and the sample-positioner means to prevent contact and resulting interference therebetween during scanning. 5. The scanning microscope recited in claim 4, wherein said plate is coupled to orthogonal plungers slidably engaged by conforming guides in said support structure, said plungers being engaged by cooperatively-aligned push-pull mechanisms for sliding the plungers in and out of the guides as necessary in order to effect a desired translation of the plate to target planar coordinates. 6. The scanning microscope recited in claim 5, wherein said push-pull mechanisms consists of screw gears engaged by drivegear devices mounted on the support structure. 7. The scanning microscope recited in claim 4, further comprising apparatus for manually controlling the motion of the plate in response to a visual signal indicative of the relative position of a desired target point on the surface of the sample with respect to the tip of the probe. 8. The scanning microscope recited in claim 1, further comprising apparatus for manually controlling the motion of the sample-positioner means in response to a visual signal indicative of the relative position of a desired target point on the surface of the sample with respect to the tip of the probe. 9. The scanning microscope recited in claim 1, further comprising apparatus for automatically controlling the motion of the sample-positioner means to provide said gap between the stand-off and the sample-positioner means in response to a signal indicative of a desired target point on the surface of the sample having been placed in alignment with the tip of said probe. 10. The scanning microscope recited in claim 1, wherein said stand-off further comprises a magnet and said free end of the scanning means comprises ferromagnetic material to improve the slidable connection therewith. 11. The scanning microscope recited in claim 1, 12. A method of analyzing a sample having an exposed surface by scanning microscopy, comprising the following steps: |
045227829 | summary | FIELD OF THE INVENTION The invention concerns a fuel assembly for a nuclear reactor constituted by a bundle of parallel fuel rods whose spacing is maintained by spacer grids which are transverse with respect to the rods and the support tubes which replace some fuel rods. BACKGROUND The support tubes or guide tubes, which are longer than the fuel rods, in cooperation with two transverse end plates, assure retention of the grids, the rigidity of the assembly, guiding of the control rods and taking up of axial forces. In the parts of the guide tubes between the upper tip and the upper end grid and also between the lower tip and the lower end grid, axial forces are transmitted only by the guide tubes, while in the whole of the general region of the assembly located between the end grids, axial force is also transmitted by the rods. The core in pressurized water nuclear reactors, for example, is constituted by such very long, square-sectioned assemblies. The end plates of such assemblies are generally made of stainless steel, while the spacer grids disposed at regular intervals along the length of the assembly are generally made of a nickel alloy or a zirconium alloy. In practice, these materials have adequate mechanical and physical properties for constituting these parts of the assembly. In known embodiments, provision is made to fix all the guide tubes to each spacer grid so that each guide tube shares in the taking up of axial forces. When the guide tubes and the spacer grids are stainless steel, the various parts, i.e., the end plates, the spacer grids and the guide tubes, constituting the framework of the assembly inside which the fuel rods are positioned, are easily connected by electrical welding. Materials such as stainless steel or nickel alloys, however, are materials with high neutron capture cross sections, so that the performance of fuel assemblies including a significant mass of these materials is proportionately reduced. The use of guide tubes made of a material with a low neutron capture cross section, such as a zirconium alloy, has therefore been proposed, e.g., in French Pat. No. 2,049,108. The zirconium alloys used, however, cannot be welded to stainless steel and the nickel alloys respectively constituting the end plates and the assembly grids. The French patent therefore proposes using intermediate connecting apparatuses welded onto the end plates and onto the grids and effecting mechanical attachment between the guide tubes made of zirconium alloy (Zircaloy) and these intermediate connecting parts. The latter can be sleeves made of stainless steel, for example, into which the tubes made of zirconium alloy are introduced and fitted securely. Such an assembly includes a smaller mass of material with a high neutron capture cross section in relation to previously known assemblies, but many complex operations are involved in manufacturing it, since stainless steel sleeves must be welded into each of the cells of the spacer grids inside which the guide tubes pass, guide tubes must be introduced into these sleeves and then guide tube must be fixed in the sleeve by introducing tooling inside the guide tubes. In addition, under irradiation, zirconium alloy tubes enlarge and expand less than the steel parts generally used for end parts, so that provision must be made in designing the assembly for this possibility of differential lengthening of the assembly in service. The object of the invention is therefore to propose a fuel assembly for a nuclear reactor constituted by a bundle of parallel fuel rods whose spacing is maintained by spacer grids which are transverse with respect to the rods, two traverse end plates and supporting guide tubes which replace some fuel rods, and which are longer than fuel rods, rigidly fixed to the end plates and to the spacer grids, assuring retention of the grids, rigidity of the assembly, and taking up of axial forces as well as guiding of the control rods of the reactor, the end plates being made of a material with a high neutron capture cross section, such as stainless steel, this fuel assembly having a structure which allows easier and cheaper manufacture and assures adequate resistance to axial forces. SUMMARY OF THE INVENTION To achieve this object, the fuel assembly according to the invention comprises at least two supporting guide tubes made of a material which is metallurgically compatible with that of the spacer grids, and other guide tubes constituting the major part of the guide tube of the assembly made of a material with a low neutron capture cross section, being secured only to one of the end plates, and simply engaged in the cells of the spacer grids so as to be movable with respect to these under the effect of expansion, and insuring only the guiding of the control rods. According to a first embodiment of the invention, the spacer grids and the supporting guide tubes fixed to these grids are made of the same material, with a high neutron capture cross section, as the end plates, for example of stainless steel. According to a second embodiment of the invention, the spacer grids and the at least two supporting guide tubes fixed to these grids are made of a material with a low neutron capture cross section, for example of Zirconium alloy, and the at least two supporting guide tubes are fixed to the end plates by mechanical connection, for example by screwing. Thus, though all the guide tubes serve to guide the control rods, only some of them assure the rigidity of the assembly and taking up of axial forces. |
050230433 | abstract | An actively cooled device, such as a heat shield for a fusion reactor, includes elements (10) of a heat-resistant material, particularly graphite, each being provided with at least one recess having a circular cross section, particularly a groove (14), into which a cooling pipe (16) conducting a coolant is brazed directly and with surface contact. This produces an effective dissipation of heat from the elements and a more durable connection of the elements with the cooling pipe or pipes. |
041585992 | description | DETAILED DESCRIPTION OF THE PRACTICE OF THIS INVENTION The apparatus shown in the drawings includes a nuclear reactor 11 which is mounted in a pit 13. This apparatus also includes a lifting device 15, typically a crane, for lifting and moving the upper package of the reactor 11. The crane 15 includes the usual drives (not shown) necessary for operation and is preferably movable on a track (not shown). There is also a pit 17 for replacement apparatus. The cart 19 containing the replacement fuel before refueling and the spent fuel after refueling is temporarily retained in pit 17. There is a spent fuel pit 18 connected to the pit 17 through a gate 20. During refueling the pit 17 and the pit 13 are filled with water. In the event that there is a necessity to effect repairs the gate 20 is closed and the water is pumped from pit 17. A crane 23 is provided for carrying out the fuel replacement operation. This crane is movable on a track 25 and has the necessity movable mast 27 to hold and move the fuel assemblies. The reactor 11 includes a lower package 31 and an upper package 33. For the details of the structure the reactor 11 and the above-listed U.S. patents (particularly patent No. 3,607,629) are referred to. For the purpose of the invention it is necessary only to describe briefly the features shown in FIGS. 1 and 2. The lower package 31 includes the body 35 of a pressure vessel. Within the body 35 there is a fuel core 37 composed of fuel assemblies (FIG. 2). Conductors 39 and 41 are connected peripherally to the body 35 for transmitting fluid between steam generators 43 and the pressure vessel and between pumps 45 and the pressure vessel. The upper package 33 includes the head 51 of the pressure vessel. The housings (not shown) for the control rid drive mechanism (not shown) are joined pressure-tight to the head 51. The driving pistons (not shown) the drive shafts (not shown), and the control rods (not shown) extend from the housings (not shown) through the head into the core 37. These drives, shafts, and control rods are held magnetically in the retracted position during refueling (see U.S. Pat. No. 3,766,006) and are removable as part of the upper package 33. The upper package also includes the missile shield 53. The shield 53 has a lifting lug 55 to which the lifting device 15 is connected for removal of the upper package 33. The casing 57 includng the control rod drive mechanisms (not shown) has a ring 59 on which a cable tray 61 is pivoted. The cables 65 are connected to the usual cabinets 63 in a compartment 67. Because the tray 61 is pivoted and the cables 65 hang loosely from the tray, the upper package may be moved without disconnecting the cables, (U.S. Pat. No. 3,836,430). The head 51 is secured pressure-tight to the body 35 of the pressure vessel by studs 71. Each stud 71 is provided with a hydraulically operable detensioner 73 for rapid movement of the studs. Typically, there are 26 large studs rather than 52 smaller studs. The studs 71 and detensioners 73 and head 51 are enclosed in a thermal insulating housing 75. The pit 13 is surrounded by an outer concrete building 81 which serves as a radioactive shield. Within building 81 there is a containment pressure shell 83, typically of steel, which has a base 85 that lines the base 87 of the building 81. The crane 23 and its tracks 25 are disposed within the shell 83. Within the shell 83 there is a concrete inner structure 91 in which the reactor 11 and its associated apparatus such as the steam generator 43, the pumps 45 and fluid conductors 39 and 41 are mounted. The reactor 11 is disposed in the pit 13 which is part of this structure 91. The vessel 35 has linings 95 of thermally insulating material separated from the wall 93 of the pit by an air space (not shown). The conductors 39 and 41 pass through openings in this structure, and have linings 97 of thermally insulating material also separated from the walls 97 by air spaces. The stem generators 43 and the pumps 45 are supported from the base 100 of the structure. The components 39, 41, 43 and 45 and other like components are accessible or may be viewed through inspection hatches or openings 101 and 103 (see FIG. 1). The operating deck above the reactor is shielded from the generators 43 by arcuate concrete shields 105 which span walls 107 secured to an arcuate concrete structure 109 extending from the shell 83. Openings 104 are provided for venting the pit 13. The base 99 is provided with a circular receptacle 111 to which the upper package 33 may be removed during refueling. There may also be receptacles 113 and 115 for storing the head 51 and the upper or lower internals (not shown) where necesssary. There is also a pressurizer 116 for maintaining the pressure during operation of the reactor 11. The inner structure 91 is provided with a hatch cover 117 of steel and concrete shown raised in FIG. 2. This cover 117 is mounted on hinges 119 and when raised affords access to the reactor 11 and its associated apparatus. When lowered, during normal operation of the reactor 11, the cover 117 affords shielding protection. Within the shell 83 there is a cavity 121 where the cart 19 with replacement fuel assemblies (or, after refueling, with the spent assemblies) is disposed during refueling. This cavity 121 is enclosed by concrete walls 123 and is separated from the inside of the inner structure 91 by a removable gate 125. The refueling pit 17 is a cavity in the wall 131 of the spent fuel storage pit 18. This wall 131 on one side abuts the adjacent wall 133 of the concrete shield 81. There is a passage 135 extending through the walls 131, 133 and 123. The passage 135 has a cylindrical metal wall 136. It is closed on the side of the pit 17 by a gate 137 which is slidable between open and closed position. On the side of the cavity 121 this passage is closed by a cover 139 which is bolted to a flange 141 of the wall defining the passage 135. OPERATION The refueling is carried out under water to protect against radioactivity. It is essential that the upper internals be maintained under water as they are raised as part of the upper package. The water level is then raised as the upper package is raised. But, the water level must be maintained below the detensioners 73, the cable tray 61 and the cable 65. Initially, the inside of the pit 13 is dry. At the start of the refueling the fuel cart 19 is disposed under water in pit 17 with replacement assemblies 151. The boron concentration is now appropriately raised to the refueling magnitude (2500) p.p.m. for annual cycle and 1440 p.p.m. or less for shorter cycles. The control rod drive mechanisms (not shown) are operated to remove the rods (not shown) to the retracted position (U.S. Pat. No. 3,766,006). The detensioners 73 are now operated and the studs 71 tilted out of locking position. The upper package is now lifted to position A (FIG. 1) and moved to the receptacle 111 in position B. At this point, or before, the level of the water in the pit 13 is raised. The level of the rising water during this operation is always maintained above the upper internals in the rising package but below the stud detensioners, cable tray 61 and cables 65. Gate 137 is then raised and cart 19 moved into cavity 121. Gate 125 is now raised and refueling carried out by crane 23. The replacement assemblies 151 one-by-one, replace spent assemblies; the spent assemblies being placed in the cart 19. After the replacement of fuel assemblies has been completed the cart 19 of spent assemblies are removed to the pit 17 whence the assemblies may be transferred under water into pit 18. The upper package is replaced as the water level is gradually lowered. The studs are secured sealing the head 51 to the body 35. The shield 53 is lowered to operating position and the control rods reinserted in core 37. The boron concentration of the water in the reactor vessel is now reduced to operating magnitude (1200 p.p.m. for annular refueling cycle, 650 for semiannual and 350 for three-months). The reactor 11 may now be set into normal operation. In lieu of refueling from a cart 19 of assemblies, the assemblies may be brought one-by-one into the containment 13 as they are at present. FIG. 3 shows the relationship between refueling cycle in months, plotted horizontally, and fuel cost saving in dollars per kilowatt, plotted vertically. Curves are plotted for three parameters of refueling time. FIG. 3 shows that the fuel savings is a maximum for a three-months cycle and refueling time of three days. While a method of practicing this invention has been disclosed herein, many modifications thereof are feasible. This invention then is not to be restricted except insofar as is necessitated by the spirit of the prior art. |
abstract | Energy sources and methods for curing in an imprint lithography system are described. The energy sources may include one or more energy elements positioned outside of the viewing range of an imaging unit monitoring elements of the imprint lithography system. Each energy source is configured to provide energy along a path to solidify polymerizable material on a substrate. |
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046997610 | summary | CROSS REFERENCE TO RELATED APPLICATIONS Reference is hereby made to the following copending applications dealing with related subject matter and assigned to the assignee of the present invention: 1. "Reconstitutable Nuclear Reactor Fuel Assembly With Unitary Removable Top Nozzle Subassembly" by John M. Shallenberger, assigned U.S. Ser. No. 673,681 and filed Nov. 20, 1984, a continuation-in-part of copending U.S. patent application Ser. No. 457,790 filed Jan. 13, 1983. 2. "Improved Removable Top Nozzle Subassembly For A Reconstitutable Nuclear Fuel Assembly" by John F. Wilson et al, assigned U.S. Ser. No. 701,049 and filed Feb. 12, 1985. BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to reconstitutable fuel assemblies for nuclear reactors and, more particularly, is concerned with an integral reusable locking arrangement for a removable top nozzle which requires no replacement components nor remote manipulations and thereby reduces the complexity of the fuel assembly reconstitution operation. 2. Description of the Prior Art Conventional designs of fuel assemblies include a mulitiplicity of fuel rods held in an organized array by grids spaced along the fuel assembly length. The grids are attached to a plurality of control rod guide thimbles. Top and bottom nozzles on opposite ends of the fuel assembly are secured to the control rod guide thimbles which extend above and below the opposite ends of the fuel rods. At the top end of the fuel assembly, the guide thimbles are attached in openings provided in the top nozzle. Conventional fuel assemblies also have employed a fuel assembly hold-down device to prevent the force of the upward coolant flow from lifting a fuel assembly into damaging contact with the upper core support plate of the reactor, while allowing for changes in fuel assembly length due to core induced thermal expansion and the like. Such hold-down devices have included the use of springs surrounding the guide thimbles, such as seen in U.S. Pats. 3,770,583 and 3,814,667 to Klumb et al and 4,269,661 to Kmonk et al, and in the first patent application cross-referenced above. Due to occasional failure of some fuel rods during normal reactor operation and in view of the high costs associated with replacing fuel assemblies containing failed fuel rods, the trend is currently toward making fuel assemblies reconstitutable in order to minimize operating and maintenance expenses. Conventional reconstitutable fuel assemblies incorporate design features arranged to permit the removal and replacement of individual failed fuel rods. Reconstitution has been made possible by providing a fuel assembly with a removable to nozzle. The top nozzle is mechanically fastened usually by a threaded arrangement to the upper end of each control rod guide thimble, and the top nozzle can be removed remotely from an irradiated fuel assembly while it is still submerged in a neutron-absorbing liquid. Once removal and replacement of the failed fuel rods have been carried out on the irradiated fuel assembly submerged at a work station and after the top nozzle has been remounted on the guide thimbles of the fuel assembly, the reconstituted assembly can then be reinserted into the reactor core and used until the end of its useful life. One recently proposed design for a reconstitutable fuel assembly top nozzle is described and illustrated in the second patent application crossreferenced above. The proposed top nozzle overcomes certain problems associated with the removable top nozzle designs of the above-cited U. S. patents and patent application. Specifically, it includes improved structures which eliminate relative moving contact or sliding engagement between the upper core support plate and the hold-down structure of the top nozzle while providing removable mounting of the top nozzle as a unitary subassembly on the guide thimbles of the reconstitutable fuel assembly as well as desired alignment of the fuel assembly with the upper core plate. Relative motion between the upper core plate and alignment sleeves of the top nozzle still takes place but without damaging contact with one another. Also, deleterious affects on the hold-down springs of coolant cross-flow between fuel assemblies is substantially eliminated in the proposed top nozzle by the incorporation of an enclosure wall about the perimeter of the top nozzle which protects the springs from flow induced vibration. The proposed top nozzle includes an upper hold-down plate which compresses a plurality of holddown springs, preferably four in number, against a lower adapter plate. Lugs associated with the upper hold-down plate and the enclosure associated with the lower adapter plate are slidably coupled together via shear pins which restrict their relative movement away from each other but allow movement toward each other. Elongated alignment sleeves at their upper ends extend through passageways in the upper hold-down plate and at their lower ends are releasably threaded to the upper ends of the guide thimbles which extend upwardly through holes in the lower adapter plate. In such arrangement, when the lower ends of the alignment sleeves are tightened against the adapter plate, they cooperate with lower retainers on the guide thimbles below the adapter plate to mount the top nozzle via the adapter plate on the guide thimbles. Conversely, for removal of the top nozzle, the alignment sleeves are unthreaded from the guide thimbles. The sleeves remain captured between the upper hold-down and lower adapter plates, as also do the respective hold-down springs which encompass the sleeves, after the sleeves are threadably disconnected from the guide thimbles and the top nozzle is removed as a unit from the fuel assembly. While the proposed removable top nozzle design of the second cross-referenced patent application, as just briefly described, has proven to be a highly satisfactory solution to the problems previously existing, it has been recently recognized that it is desirable to provide some additional means to ensure that the top nozzle remains attached to the guide thimbles during in-core operation subsequent to reconstitution. Heretofore, locking techniques employed to prevent in-service loosening of the top nozzle attachment have taken several forms, such as; swagged locking cups on cap screws, swagged expansions of tubes into collars, welded lock pins, etc. Regardless of the locking technique employed, one similarity exists between them all: they require handling of loose parts for disassembly and replacement of the locking feature (i.e., a new lock pin, locking cup, etc.) for both top nozzle removal and replacement operations and elaborate remote tooling to accomplish the removal and relocking tasks. Consequently, a need exists for a different approach to locking the removable top nozzle to the guide thimbles of the reconstitutable fue assembly, one with the objective of accommodating the beneficial design features of the top nozzle disclosed in the second cross-referenced application while avoiding the complicated locking techniques employed in the past. SUMMARY OF THE INVENTION The present invention provides an integral reusable locking arrangement for the removable top nozzle designed to satisfy the aforementioned needs. While the threaded joints connecting the top nozzle to the guide thimbles are still employed, a locking arrangement is integrally incorporated which is reusable and requires no special tooling to actuate it. Basically, as before, the alignment sleeves are just unthreaded from and rethreaded onto the upper ends of the guide thimbles to remove and replace the top nozzle. There is no need for replacement components nor additional remote operations such as the making of expansions or swages as commonly used in previously existing designs. As a result, the complexity of the fuel assembly reconstitution operation is greatly reduced. Accordingly, the present invention is set forth in a reconstitutable fuel assembly having at least one control rod guide thimble and a top nozzle. The control rod guide thimble includes an upper end portion having an annular externally threaded section. The top nozzle includes at least one tubular alignment sleeve having a lower annular internally threaded section. The alignment sleeve is rotatably movable relative to the guide thimble upper end portion between lowered and raised positions for threading and unthreading its internally threaded section onto and from the externally threaded section of the guide thimble upper end portion for attaching and detaching the top nozzle onto and from the guide thimble. The present invention provides an integral reusable locking arrangement for the top nozzle comprising: (a) inner means mounted on the guide thimble upper end portion above the externally threaded section thereon; and (b) outer means mounted on the alignment sleeve above the internally threaded section thereon: (c) the inner and outer means being configured to coact with one another as the internally threaded section of the alignment sleeve is threaded on the externally threaded section of the guide thimble upper end portion when attaching the top nozzle to the guide thimble and to provide a locking force which must be overcome in order to unthread the internally threaded section of the alignment sleeve from the externally threaded section of the guide thimble upper end portion for detaching the top nozzle from the guide thimble. The locking force takes the form of a constant torsional drag produced between the inner and outer means as the alignment sleeve is rotatably moved relative to the guide thimble upper end portion between the lowered and raised positions. Additionally, the outer means on the alignment sleeve and the inner means on the guide thimble upper end portion are axially displaced from one another when the internally threaded section on the alignment sleeve is initially rotatably moved into threaded engagement with the externally threaded section on the guide thimble upper end portion whereby the mechanical advantage produced by threading the internally threaded section of the alignment sleeve onto the externally threaded section of the guide thimble upper end portion can be used to overcome the locking force as the alignment sleeve is threaded onto the guide thimble upper end portion. More particular, the inner means takes the form of a thin-walled tubular section on the guide thimble upper end portion having a first external diametric size and an enlarged region defined thereon of a second external diametric size which is greater than the first external diametric size of the remainder of the tubular section. The outer means takes the form of an axial section on the alignment sleeve having a third internal diametric size which is greater than the first external diametric size of the thin-walled tubular section on the guide thimble upper end portion but less than the second external diametric size of the enlarged region on the tubular section, whereby rotational movement of the alignment sleeve relative to the guide thimble upper end portion from the raised toward the lowered position causes interference contact of the axial section of the alignment sleeve with the tubular section enlarged region of the guide thimble upper end portion so as to produce the locking force. Specifically, the enlarged region takes the form of a circumferential protrusion on the tubular section. |
description | The present application is a U.S. national stage application under 35 U.S.C. § 371 of PCT/US2013/066777 filed Oct. 25, 2013, which claims the benefit of U.S. Provisional Patent Application No. 61/718,309 filed Oct. 25, 2012, the entireties of which are herein by reference. The present invention relates to nuclear power generation systems, and more particularly to a steam supply system and related power cycle including natural thermally induced gravity flow of primary coolant through a reactor and steam generator. Commercial nuclear power generation plants in use as well as those thus far submitted for licensing approval in the United States and abroad are characterized by the following common features with respect of their power generation cycle (e.g. Rankine cycle). First, the motive steam produced which is used to drive the steam turbine is produced at a moderately high pressure with little or no superheat. Second, multiple trains of feedwater heaters are deployed to extract cycle steam at various points and pressures as the steam expands through the turbines (typically high pressure and low pressure turbine units) to pre-heat the condensate (i.e. condenses secondary coolant) before it is returned to the boiler (which may be the reactor itself or a tubular heat exchanger commonly known as the steam generator). Finally, in addition to the string of feedwater heaters, the cycle steam is also generally mechanically dewatered and reheated by the boiler steam (without mixing) in one or two stages in a rather large piece of equipment known as the moisture separator reheater or MSR. The array of steam turbines, feedwater heaters, MSR, and associated piping, valves and instrumentation, typical of modern nuclear plants, are also a known source of reduction in the plant's reliability and increase in maintenance costs. Although the foregoing equipment mentioned may add perhaps one to two percent to the plant's thermodynamic efficiency, the concomitant increase in complexity, capital cost, and detriment to the plant's availability may be quite significant. In particular, the complexity of the power cycle detracts from the use of the nuclear plant as a variable power supply source for cyclical “load following” operation. The present disclosure provides an improved nuclear steam supply system and optimized Rankine cycle which includes natural thermally induced gravity flow to circulate primary coolant through the steam generating and reactor vessels combined with a low pressure steam supply to the turbine-generator set, thereby advantageously eliminating the need for a high pressure turbine. In addition, embodiments of the present invention may include a single feedwater heater in the secondary coolant flow or piping loop external to the reactor vessel. This arrangement produces minimal heating of the liquid phase secondary coolant upstream and outside of the steam generator in the balance of plant, thereby maximizing the heat transfer rate from the primary coolant to the secondary coolant within the steam generating vessel. Advantageously, this results in a larger temperature differential initially between the primary coolant and secondary coolant in the steam generator because the bulk of the secondary coolant heating to produce steam for driving the low pressure turbine is accomplished within the steam generating vessel. This in turn enhances and increases the thermo-siphon effect mechanism of the natural convection-based gravity-driven primary coolant circulation system and rate of primary coolant flow, thereby advantageously eliminating the need for primary coolant pumps, as further described herein. A natural convection driven reactor primary coolant loop coupled with an optimized simple steam cycle as disclosed herein is advantageously well adapted to load following operation, which heretofore has been impractical and problematic for modern large nuclear plants. In one embodiment, a nuclear power generation system includes a steam generator including a steam generating vessel fluidly coupled to a reactor vessel having an internal cavity, a reactor core comprising nuclear fuel disposed within the internal cavity and operable to heat a primary coolant, a primary coolant flow loop formed between the reactor vessel and the steam generating vessel, the primary coolant flow loop being configured and operable to circulate primary coolant through the reactor vessel and steam generating vessel via thermally induced gravity flow, and a secondary coolant flow loop formed between the steam generating vessel and a low pressure turbine. The secondary coolant flow loop is configured and operable to circulate secondary coolant through the steam generating vessel in which the primary coolant heats and converts the secondary coolant from liquid to steam, the steam flowing through the secondary coolant flow loop to the low pressure steam turbine. A temperature differential between primary coolant leaving the steam generating vessel and secondary coolant entering the steam generating vessel is sufficient to induce natural thermally driven gravity circulation of the primary coolant through the primary coolant flow loop. In another embodiment, a nuclear power generation system includes a steam generator including a steam generating vessel fluidly coupled to a reactor vessel having an internal cavity, a reactor core comprising nuclear fuel disposed within the internal cavity and operable to heat a primary coolant, a primary coolant flow loop formed between the reactor vessel and the steam generating vessel, the primary coolant flow loop being configured and operable to circulate primary coolant through the reactor vessel and steam generating vessel via thermally induced gravity flow, and a secondary coolant flow loop formed between the steam generating vessel and a single steam turbine, the secondary coolant flow loop being configured and operable to circulate secondary coolant through the steam generating vessel in which the primary coolant heats and converts the secondary coolant from liquid to steam. The steam flows through the secondary coolant flow loop to an inlet on the low pressure steam turbine. A condenser is disposed in the secondary coolant flow loop between an outlet from the low pressure turbine and the steam generating vessel, the condenser configured to cool and condense steam exiting the low pressure turbine thereby converting the secondary coolant from steam to liquid. A single feedwater heater is disposed in the secondary coolant flow loop between the condenser and the steam generating vessel, the feedwater heater configured to receive and heat the liquid secondary coolant from the condenser, wherein the secondary coolant flows directly from the feedwater heater into the steam generating vessel without any intervening feedwater heaters between the condenser and the steam generating vessel. A temperature differential between the primary coolant leaving the steam generating vessel and secondary coolant entering the steam generating vessel is sufficient to induce natural thermally driven gravity circulation of the primary coolant through the primary coolant flow loop. A method for inducing thermally driven gravity flow of primary coolant through a nuclear reactor is provided. The method includes providing a steam generating vessel fluidly coupled to a reactor vessel housing a nuclear fuel core which heats a primary coolant; circulating the primary coolant through a primary coolant flow loop formed between the steam generating vessel and reactor vessel, the primary coolant entering the steam generating vessel from the reactor vessel at a first temperature and exiting the steam generating vessel at a second temperature lower than the first temperature; heating a secondary coolant in the steam generating vessel using the primary coolant which converts the secondary coolant from a liquid entering the steam generating vessel to steam exiting the steam generating vessel, the secondary coolant entering the steam generating vessel at a third temperature and exiting the steam generating vessel at a fourth temperature higher than the third temperature; circulating the secondary coolant through a secondary coolant flow loop having an external portion outside to the steam generating vessel; expanding the steam in a low pressure steam turbine for producing electric power; condensing the steam in a surface condenser to convert the secondary coolant from steam back into liquid form; heating the liquid secondary coolant received from condenser in a feedwater heater to the third temperature using fluid extracted from the turbine; and flowing the heated liquid secondary coolant from the feedwater heater directly to the steam generating vessel; wherein the temperature differential between the second temperature of the primary coolant and the third temperature of the secondary coolant is selected to induce natural thermally driven gravity circulation of the primary coolant through the primary coolant flow loop. 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. The features and benefits of the invention are illustrated and described herein by reference to non-limiting exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description, and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Referring first to FIG. 1, a nuclear steam supply system 100 is illustrated in accordance with an embodiment of the present invention. Although described herein as being a nuclear steam supply system, the system may also be generally referred to herein as a “steam supply system” for brevity. The inventive nuclear steam supply system 100 is typically used in a nuclear pressurized water reactor. Of course, the nuclear steam supply system 100 can have uses other than for nuclear pressurized water reactors as can be appreciated. In one embodiment, the nuclear steam supply system 100 generally comprises a reactor vessel 200 and a steam generating vessel 300 fluidly coupled to the reactor vessel. During normal operation of the nuclear steam supply system 100, a primary coolant flows through a primary coolant flow loop 190 within the reactor vessel 200 and the steam generating vessel 300. This primary coolant loop 190 is depicted with arrows in FIG. 1. Specifically, the primary coolant flows upwardly through a riser column 224 in the reactor vessel. 200, from the reactor vessel 200 to the steam generating vessel 300 through a fluid coupling 270, upwardly through a riser pipe 337 in the steam generating vessel 300 to a top of the steam generating vessel 300 (i.e., to a pressurizer 380), and then downwardly through tubes 332 (see FIGS. 3 and 4) in a tube side 304 of the steam generating vessel 300, from the steam generating vessel 300 to the reactor vessel 200 through the fluid coupling 270, downwardly through a downcomer 222 of the reactor vessel 200, and then back from the downcomer 222 of the reactor vessel 200 to the riser column 224 of the reactor vessel 200. The primary coolant continues to flow along this primary coolant loop 190 as desired without the use of any pumps during normal operation of the nuclear steam supply system 100. During normal operation of the nuclear steam supply system 100, the primary coolant has an extremely high temperature due to its flowing through and heating by the reactor core 230. Specifically, nuclear fuel in the reactor vessel 200 engages in the fission chain reaction, which produces heat and heats the primary coolant as the primary coolant flows through the reactor core of the reactor vessel 200. This heated primary coolant is used to phase-change a secondary coolant from a liquid into steam used to drive the turbine-generator set, as discussed below. While the primary coolant is flowing through the primary coolant loop 190 during normal operation, the secondary coolant is flowing through a second coolant loop. Specifically, the secondary coolant is introduced into the shell side 305 (FIGS. 3 and 4) of the steam generating vessel 300 at the secondary coolant in location indicated in FIG. 1. The secondary coolant then flows through the shell side 305 (FIGS. 3 and 4) of the steam generating vessel 300 where it is heated by heat transfer from the primary coolant. The secondary coolant is converted into steam due to the heat transfer, and the steam flows from the steam generating vessel 300 to a turbine 900 as indicated in FIG. 1. The turbine 900 drives an electric generator 910 which is connected to the electrical grid for power distribution. The steam then travels from the turbine 900 to a condenser (not illustrated) whereby the steam is cooled down and condensed to form condensate. Thus, the condenser converts the steam back to a liquid condensate (i.e., the secondary coolant) so that it can be pumped back into the steam generator 300 at the secondary coolant inlet location and repeat its flow through the flow path discussed above and be converted back to steam. In certain embodiments both the primary coolant and the secondary coolant may be water, such as demineralized or borated water. However, the invention is not to be so limited and other liquids or fluids can be used in certain other embodiments, the invention not being limited to the material of the primary and secondary coolants unless so claimed. The primary coolant continues to flow through the primary coolant loop and the secondary coolant continues to flow in the second coolant loop during normal operation of the nuclear steam supply system 100. The general provision and operation of the convective thermal gravity-driven nuclear steam supply system 100, as well as details of the associated components are described in detail in co-pending International Application No. PCT/US13/38289 filed Apr. 25, 2013, the entirety of which is incorporated herein by reference. Referring to FIGS. 1-4, the general details of the components and the operation of the nuclear steam supply system 100, and specifically of the reactor vessel 200 and the steam generating vessel 300, will be described. In the exemplified embodiment, the reactor vessel 200 and the steam generating vessel 300 are vertically elongated and separate components which hydraulically are closely coupled, but are discrete vessels in themselves that are thermally isolated except for the exchange of primary coolant (i.e. reactor coolant) flowing between the vessels in the fluid coupling 270 of the primary coolant loop 190 as discussed above. In one non-limiting embodiment, each of the reactor vessel 200 and the steam generating vessel 300 may be made of a corrosion resistant metal such as stainless steel, although other materials of construction are possible. Referring to FIGS. 1 and 2 concurrently, the reactor vessel 200 will be further described. The reactor vessel 200 in one non-limiting embodiment is an ASME code Section III, Class 1 thick-walled cylindrical pressure vessel comprised of a cylindrical sidewall shell 201 with an integrally welded hemispherical bottom head 203 and a removable hemispherical top head 202. The shell 201 defines an internal cavity 208 configured for holding the reactor core which comprises the nuclear fuel. Specifically, the reactor vessel 200 includes a cylindrical reactor shroud 220 which contains the reactor core defined by a fuel cartridge 230 (i.e., nuclear fuel). The reactor shroud 220 transversely divides the shell portion of the reactor vessel into two concentrically arranged spaces: (1) an outer annulus 221 defining the annular downcomer 222 for primary coolant entering the reactor vessel which is formed between the outer surface of the reactor shroud 220 and an inner surface 206 of the shell 201; and (2) a passageway 223 defining the riser column 224 for the primary coolant leaving the reactor vessel 200 heated by fission in the reactor core. The reactor shroud 220 is elongated and extends in an axial direction along a vertical axis A-A of the reactor vessel 200. The reactor shroud 220 includes an open bottom end 225 and a closed top end 226. In one embodiment, the open bottom end 225 of the reactor shroud 220 is vertically spaced apart by a distance from the bottom head 203 of the reactor vessel 200 thereby forming a bottom flow plenum 228 between the bottom end 225 of the reactor shroud 220 and the bottom head 203 of the reactor vessel 200. As will be discussed in more detail below, during flow of the primary coolant through the primary coolant loop 190, the bottom flow plenum 228 collects the primary coolant from the annular downcomer 222 and directs the primary coolant flow into the inlet of the riser column 224 formed by the open bottom end 225 of the reactor shroud 220. In certain embodiments, the reactor shroud 220 is a double-walled cylinder which may be made of a corrosion resistant material, such as without limitation stainless steel. This double-wall construction of the reactor shroud 220 forms an insulated structure designed to retard the flow of heat across it and forms a smooth vertical riser column 224 for upward flow of the primary coolant heated by the fission in the fuel cartridge 230 (“core”), which is preferably located at the bottom extremity of the shroud 220 in one embodiment as shown in FIG. 2. The vertical space above the fuel cartridge 230 in the reactor shroud 220 may also contain interconnected control rod segments along with a set of “non-segmental baffles” that serve to protect them from flow induced vibration during reactor operations. The reactor shroud 220 is laterally supported by the reactor vessel by support members 250 to prevent damage from mechanical vibrations that may induce failure from metal fatigue. In certain embodiments, the fuel cartridge 230 is a unitary autonomous structure containing upright fuel assemblies, and is situated in a region of the reactor vessel 200 that is spaced above the bottom head 203 so that a relatively deep plenum of water lies underneath the fuel cartridge 230. The fuel cartridge 230 is insulated by the reactor shroud 220 so that a majority of the heat generated by the fission reaction in the nuclear fuel core is used in heating the primary coolant flowing through the fuel cartridge 230 and adjoining upper portions of the riser column 224. In certain embodiments, the fuel cartridge 230 is an open cylindrical structure including cylindrically shaped sidewalls, an open top, and an open bottom to allow the primary coolant to flow upward completely through the cartridge (see directional flow arrows, described in detail above with specific reference to FIG. 1). In one embodiment, the sidewalls of the fuel cartridge 230 may be formed by multiple arcuate segments of reflectors which are joined together by suitable means. The open interior of the fuel cartridge 230 may be filled with a support grid for holding the nuclear fuel rods and for insertion of control rods into the core to control the fission reaction as needed. In the interconnecting space between the reactor vessel 200 and the steam generating vessel 300 there is a fluid coupling 270 that comprises an inner flow path 271 and an outer flow path 272 that concentrically surrounds the inner flow path 271. As will be discussed in more detail below, during flow of the primary coolant the primary coolant flows upwardly within the riser column 224 and through the inner flow path 271 of the fluid coupling 270 to flow from the reactor vessel 200 to the steam generating vessel 300. After the primary coolant gets to the top of the steam generating vessel 300, the primary coolant begins a downward flow through the steam generating vessel 300 and then flows through the outer flow path 272 from the steam generating vessel 300 and into the downcomer 222 of the reactor vessel 200. Again, this flow path will be described in more detail below. Turning now to FIGS. 1, 3, and 4, the details of the steam generating vessel 300 will be described in more detail. In certain embodiments, the steam generating vessel 300 includes a preheater section 320, a steam generator section 330, a superheater section 340 and a pressurizer 380. However, the invention is not to be so limited and one or more of the sections of the steam generating vessel 300 may be omitted in certain other embodiments. Specifically, in certain embodiments, the preheater section 320 may be omitted, or instead may be formed as an integral part of the secondary coolant inlet portion of the steam generator section. 330 without separate discrete tubesheets. In this latter arrangement, liquid secondary coolant from the feedwater heater 540 (see FIG. 6) still enters the preheater section 320 in liquid phase (thereby defining the “preheater”) and is converted to steam in an above portion of steam generator section 330 (thereby differentiating and defining the “steam generator section”). In other possible embodiments, the preheater section 320 may include separate tubesheets of its own which physically terminate and defines the limits of the preheater section. A steam bypass loop 303 may be provided on shell side of steam generating vessel 300 (see, e.g. FIG. 1) to route saturated steam from the steam generator section 330 upwards to the superheater section 340 around the intermediate tubesheet structure as shown. As discussed above, it is within the steam generator vessel 300 that the secondary coolant that is flowing through the shell side 305 of the steam generator vessel 300 is converted from a liquid phase (i.e. secondary coolant liquid inlet shown in FIG. 1) to a superheated steam phase that is sent to the turbine 900 through secondary coolant flow loop 500 (secondary coolant steam outlet shown in FIG. 1) for electricity generation via generator 910. The secondary coolant flows in the second coolant loop 500 through the shell side of the steam generating vessel 300, out to the turbine 900, from the turbine 900 to a condenser 520, through a single feedwater heater 540, and then back into the shell side of the steam generating vessel 300, as shown in FIG. 6 and further described herein. In the exemplified embodiment, each of the preheater 320, the steam generator 330, and the superheater 350 sections are tubular heat exchangers having a tube side 304 and a shell side 305. The tube side 304 of the tubular heat exchangers include a tube bundle comprising a plurality of parallel straight tubes 332 and tubesheets 333 disposed at the extremities or ends of each tube bundle that support the tubes. In the exemplified embodiment, only two tubes 332 are illustrated for clarity; however, it will be appreciated that an actual installation may include tens, hundreds, or thousands of tubes 332 within each of the sections of the steam generating vessel 300. In certain embodiments, a bottom-most one of the tubesheets 333A is located in the preheater section 320 or in the steam generator section 330. The preheater 320, steam generator 330, and superheater 350 sections form vertically stacked heat exchangers disposed within the vertically elongated steam generating vessel 300. In certain embodiments, the preheater section 320, steam generator section 330, and superheater section 350 are configured and arranged to form a parallel counter-flow type heat exchanger arrangement in which the secondary coolant (Rankine cycle) flows upwards on the shell side in an opposite, but parallel direction to the downwards flowing primary coolant on the tube side (see FIGS. 1, 3, and 4). The primary coolant, however, flows upwards in riser pipe 337 in the same direction and parallel to the flowing secondary coolant. Specifically, the arrows labeled A indicate the flow direction of the primary coolant through the riser pipe 337 that is positioned within the steam generating vessel 300, the arrows labeled B indicate the flow direction of the primary coolant through the tubes 332 of the steam generating vessel 300, and the arrows labeled C indicate the flow direction of the secondary coolant through the shell side 305 of the steam generating vessel 300. The trio of the foregoing tubular heat exchangers (i.e. preheater, steam generator, and superheater) are hydraulically connected in series on both the tube side 304 (primary coolant) and the shell side 305 (the secondary coolant forming the working fluid of the Rankine Power Cycle which changes phase from liquid to superheated, gas in the steam generating vessel 300). It should be noted that the portion of primary coolant flowing upwards in riser pipe 337 in the same direction and parallel to the flowing secondary coolant in the steam generating vessel 300 prior to entering the tubes 332 in the pressurizer 380 may contribute slightly to heating the secondary coolant. However, the primary heat transfer occurs across the tube 332 walls between the hot primary coolant flowing inside tubes 332 and secondary coolant flowing outside the tubes. In the exemplified embodiment, the steam generating vessel 300 includes a top 310, a bottom 311, an axially extending cylindrical shell 312, and the internal riser pipe 337 which is concentrically aligned with the shell 312 and in the exemplified embodiment lies on a centerline C-C of the steam generating vessel 300. The tubes 332 are circumferentially arranged around the outside of the riser pipe 337 between the riser pipe 337 and the shell 312 in sections of the steam generating vessel 300 which include the preheater 320, the steam generator 330, and the superheater 350. In one embodiment, the riser pipe 337 extends completely through all of the tubesheets 333 associated with the preheater 320, the steam generator 330, and the superheater 350 from the top of the steam generating vessel 300 to the bottom to form a part of the continuous primary coolant loop 190 between the reactor vessel 200 and the steam generating vessel 300 all the way to the pressurizer 380. The fluid coupling 270 includes an inner flowpath 371 and an outer flowpath 372 on the steam generating vessel 300 side of the fluid coupling 270. The inner flowpath 371 is fluidly coupled to the inner flow path 271 and the outer flowpath 372 is fluidly coupled to the outer flowpath 272. Thus, via these operable couplings the steam generating vessel 300 is fluidly coupled to the reactor vessel 200 to complete the primary coolant loop 190 for flow of the primary coolant through both the reactor vessel 200 and the steam generating vessel 300. An annular space is formed between the riser pipe 337 and the shell 312, which forms a bottom plenum 338. The bottom plenum 338 collects and channels the primary coolant from the steam generating vessel 300 back to the reactor vessel 200 via the outer flow paths 272, 372. Thus, in the exemplified embodiment the primary coolant flows from the reactor vessel 200 to the steam generating vessel 300 through the inner flow paths 271, 371 and the primary coolant flows from the steam generating vessel 300 to the reactor vessel 200 through the outer flow paths 272, 372. However, the invention is not to be so limited and in other embodiments the use of the flow paths 271, 272, 371, 372 can be reversed The superheater 350 is topped by a pressurizer 380 as shown in FIGS. 1 and 4, which is in fluid communication with both the top or outlet of the riser pipe 337 and the inlet to the tubes 332 of the superheater 350. In one embodiment, the pressurizer 380 is mounted directly to the shell 312 of the steam generating vessel 300 and forms a top head 336a on the shell. In one embodiment, the pressurizer has a domed or hemispherical head and may be welded to the shell 312, or alternatively bolted in other possible embodiments. The pressurizer 380 forms an upper plenum which collects reactor primary coolant rising through riser pipe 337 and distributes the primary coolant from the riser pipe 337 to the superheater tubes 332. In certain embodiments, the pressurizer 380 includes a heating/quenching element 38. (i.e. water/steam) for pressure control of the reactor primary coolant. Shown schematically in FIG. 4, the heating/quenching element 383 is comprised of a bank of electric heaters which are installed in the pressurizer section that serve to increase the pressure by boiling some of the primary coolant and creating a steam bubble that resides at the top of the pressurizer near the head (above the liquid/gas interface 340 represented by the dashed line). A water spray column 384 is located near the top head 336a of the pressurizer 380 which sprays water into the steam bubble thereby condensing the steam and reducing the size of the steam bubble. The increase/decrease in size of the steam bubble serves to increase/decrease the pressure of the primary coolant inside the reactor coolant system. In one exemplary embodiment, a representative primary coolant pressure maintained by the pressurizer 380 and the heating/quenching element 383 may be without limitation about 2,250 psi. In alternative embodiments, as noted above, the liquid/gas interface 340 is formed between an inert gas, such as nitrogen (N2) supplied by supply tanks (not shown) connected to the pressurizer 380, and the liquid primary coolant. In one embodiment, the external surfaces of the tubes 332 may include integral fins for increasing the heat transfer rates between the hot primary coolant and secondary coolant. The superheater tube bundle is protected from erosion (i.e. by tiny water droplets that may remain entrained in the up-flowing steam) by ensuring that the steam flow is counter-flow being parallel along, rather than across, the tubes 332 in the tube bundle. In the exemplified embodiment, the nuclear steam supply system 100 including the reactor vessel 200 and the steam generating vessel 300 are housed within a containment structure or vessel 400. This ensures that in the event of a loss-of-coolant accident during start-up, all of the high energy fluids are contained within the containment boundary of the containment vessel 400. The details of the containment vessel 400 can be found in PCT/US13/42070, filed on May 21, 2013, the entirety of which is incorporated herein by reference. The Rankine power cycle components found in the secondary coolant flow loop including turbine-generator set 900-910 may be mounted outside the containment. By definition, a passively safe nuclear steam supply system as disclosed herein does not include or require any 100% primary coolant flow pumps in the primary reactor coolant loop because the flow is driven by gravity, not mechanical pumps. In a passively cooled reactor, natural circulation flow will be sustained even after the reactor shutdown control rods are fully inserted into the core. Secondary Coolant Flow Loop (Power Cycle) FIG. 6 is a schematic diagram showing the primary coolant flow loop and one embodiment of secondary coolant flow loop according to the present disclosure. The primary coolant flow loop 190 includes the reactor vessel 200 and related components already described herein. The secondary coolant flow loop 500 circulates secondary coolant via pumped flow and includes portions both inside and outside the containment vessel 400. The secondary coolant flow loop 500 includes a flow conduit (e.g. piping) for fluidly coupling components together and circulating secondary coolant in both a steam state (i.e. saturated and superheated) and a liquid state (e.g. condensate or feedwater) through the power generation equipment described herein including a turbine-generator set and auxiliary plant equipment which define a Rankine power cycle. It should be noted that FIG. 6 illustrates some representative, but non-limiting examples of optimized primary coolant and secondary coolant conditions at various points in the steam supply system 100 in one possible configuration and operating mode of the system. The invention is expressly not limited to the operating conditions shown, which represents one of many possibly nuclear plant steam supply system operating scenarios falling within the scope of the present disclosure. Referring to FIG. 6, secondary coolant flow loop 500 is generally a one-directional flow circuit including (in order along the flow path) the steam generating vessel 300 (e.g. preheat section 320, steam generator section 330, and superheater section 350), a low pressure steam turbine 900 coupled to a generator 910 (shown in FIG. 1), a condenser 520, feedwater pump 530, and a feedwater heater 540. In one embodiment, a single feedwater heater 540 is provided in the secondary coolant flow loop as further described herein. Steam leaves the steam, generating vessel 300 in a superheated state and flows through suitable steam piping of the secondary coolant flow loop 500 to an inlet on steam turbine 900 in which the steam expands and drives a plurality of rotating turbine blades. The turbine rotor is coupled to the generator 910 for producing electric current. Accordingly, in the present non-limiting embodiment being described, the secondary coolant flow loop consists of only a single low pressure turbine (e.g. no high pressure turbine) and a single feedwater heater. The steam generating vessel 300 according to the present disclosure produces steam at a relatively low pressure, but with considerable superheat for feeding a single low pressure turbine 900. Advantageously, this eliminates the need for a high pressure turbine which is generally more vulnerable to operational failures and reliability concerns due to higher operating conditions. In standard commercial nuclear plants, for example, such high pressure turbines typically may require steam at a moderately high pressure (e.g. 700 to 1000 psi) and with little or no superheat. A Rankine cycle according to the present disclosure therefore has a turbine-generator set consisting of a single low pressure steam turbine 900 coupled to the generator 910. To provide low pressure, high superheat steam to turbine 900, steam generating vessel 300 is configured and designed to produce and supply steam to low pressure turbine 900 at a pressure less than 400 psia, and in some preferred embodiments at a maximum pressure of about 350 psia. In the exemplary non-limiting power cycle illustrated in FIG. 6, the steam has a pressure of about 335 psia leaving the steam generating vessel 300 and about 325 psia at the inlet of the turbine 900 (the difference attributable to frictional pressure loss in the steam, piping between the steam generating vessel and turbine). Steam leaving the steam generating vessel 300 and entering the low pressure turbine 900 may be at least 575 degrees F. and in a superheated state. In the exemplary non-limiting power cycle illustrated in FIG. 6, the steam has a temperature of about 595 degrees F. An acceptable steam temperature range of about and including 575-615 degrees F. may be used while maintaining the low pressure steam requirements of the turbine 900. In one preferred embodiment, the steam may have a superheat of at least 170 degrees F. for optimum cycle efficiency. Due to significant initial superheat of steam entering the low pressure turbine 900, the MSRs may be completely eliminated and the multitude of feedwater heaters (e.g. 16 in a typical nuclear plant with high pressure steam turbines) may be replaced in one embodiment with a single feedwater heater 540 which draws steam and/or water (i.e. “extraction fluid”) from one or more extraction points on the low pressure turbine. The extracted steam provides the heating medium which heats the condensate/feedwater in feedwater heater 540 collected by the condenser. The feedwater heater 540 may draw steam at one or more extraction points each located on the turbine 900 to remove wet saturated steam. In one embodiment, each steam extraction point may be located on the turbine 900 at points which are at sub-atmospheric pressures (i.e. less than 14.7 psia) where air in-leakage through turbine glands is most likely to occur. In one embodiment, two sub-atmospheric extraction points may be used. A first steam extraction point E1 may be at less than 14.7 psia, and a second steam extraction point E2 may be at less than 10 psia. In one representative but no n-limiting example, the first extraction point may be about 10 psia and the second extraction point may be about 4 psia. The first steam extraction point removes generally wet saturated steam from the turbine 900 which serves to dry the cycle steam to enable it to complete the last stages of expansion in the turbine with an acceptable level of moisture to minimize droplet impingement on and wet steam erosion of the turbine blades. The second steam extraction point E2 removes extremely wet steam and/or liquid (water) secondary coolant from turbine 900. Representative but non-limiting optimal steam and/or water extraction fluid rates are noted in FIG. 6. In one embodiment, the extraction or flow rate (pounds/hour) for the lower pressure steam extraction point E2 preferably may be more than E1, and in some embodiments more than ten times E1. Extraction point E1 also has a correspondingly higher temperature, enthalpy (BTU/pound), and pressure (psia) than point E2 since the extraction fluid at E1 is withdrawn upstream of E2 from turbine 900 where the secondary coolant steam has more energy. In another possible embodiment, one above atmospheric pressure steam extraction point E1 and one sub-atmospheric steam extraction point E2 may be used. Some non-limiting representative pressures for extraction points E1, E2 may be about 20 psia and 10 psia, respectively which are optimal for this arrangement. These two extraction points serve the same foregoing purposes. Feedwater heater 540 receives steam from both extraction points E1, E2 and is suitably engineered to prevent back flow of steam and flashing of the condensate in the drain lines. Because the steam extraction occurs in the sub-atmospheric portion of the turbine in at least one of the extraction points E1, E2 where the air in-leakage is most likely to occur, the extraction steam also serves to carry the non condensibles out of the turbine space, delivering them to the feedwater heater 540 where it is segregated and exhausted using conventional vacuum apparatus. Thus, the power cycle is purged of deleterious non-condensibles which is a collateral benefit of sub-atmospheric steam extraction and feedwater heating described herein. Feedwater heater 540 may be an elongated shell and tube heat exchanger with condensate pumped from condenser 520 by feedwater pump 530 flowing on the tube side and extraction fluid (i.e. steam and/or water) from the low pressure steam turbine 900 flowing on the shell side. Although the extraction fluid from extraction points E1 and E2 is shown in FIG. 6 as flowing in a single combined flow conduit to feedwater heater 540, the extracted fluids preferably are not be combined and isolated from each other flowing in separate piping lines to the feedwater heater each having a separate inlet nozzle on the shell for introducing this heating fluid. It should be noted that the exemplary non-limiting extraction fluid conditions shown in the associated box in FIG. 6 include two sets of numbers for each parameter; the first number being for extraction point E1 fluid and the second being for the lower pressure extraction point E2 fluid. Feedwater heater 540 may be any suitable type of shell and tube heat exchanger. In one embodiment shown in FIG. 6, the feedwater heater may be a U-tube bundle design with a single channel, and tubesheet (“head”) disposed at one end of the heater. Such designs are well known in the art without undue elaboration. Essentially, the channel at the head of the heater defines an internal cavity which is internally divided by a partition plate into feedwater (secondary coolant) inlet and outlet flow boxes each in fluid communication with the single tubesheet. The U-shaped tubes have one end fluidly coupled to the tubesheet adjacent the inlet flow box and the other end fluidly coupled to the tubesheet adjacent the outlet flow box. The U-shaped bends in the tubes are disposed in the heater opposite the tubesheet and head. In other possible designs, the feedwater heater 540 may have a straight tube bundle extending between two tubesheets disposed at either end of the heater, analogous to the steam generating vessel 300 heat exchanger design described herein. Either type of heater design is suitable and may be used. The extracted and condensed secondary coolant heating medium on the shell side of feedwater heater 540 may be collected and returned to condenser 520 in one embodiment. Condenser 520 is a steam surface condenser configured to receive and condense exhaust steam (secondary coolant) from turbine 900, which is then collected in a wet well in the condenser from which feedwater pump 530 takes suction. Surface condensers of this type are well known without undue elaboration. Condenser 520 may be of any suitable design used in the industry. Feedwater pump 530 pressurizes and pumps the condensate or feedwater (terms may be used interchangeably herein to refer to the condensed liquid secondary coolant) from condenser 520 to the head of feedwater heater 540 through suitable piping. Feedwater pump 530 may be any suitable type of pump, including for example without limitation centrifugal pumps, vertical axial pumps, etc. It should be noted that because the feedwater heating of the condensate or feedwater outside of the steam generating vessel 300 in the secondary coolant loop 500 is minimal, a preponderant amount of heating of the condensate or feedwater occurs in the steam generator vessel which enables the reactor primary coolant to be cooled to much lower temperature leaving the steam generator 300 (e.g. about 375-385 degrees F. in one non-limiting example) than in a typical nuclear plant (e.g. about 500 degrees F.). Advantageously, this has the effect of boosting the natural convective thermo-siphon action in the gravity-driven reactor primary coolant loop circulation design disclosed herein. The primary coolant may be decreased in temperature by at least 200 degrees in one embodiment (i.e. entering versus leaving the steam generating vessel 300). The secondary coolant may correspondingly be increased in temperature by at least 350 degrees F. (i.e. entering versus leaving the steam generating vessel 300), and in some embodiments by at least 400 degrees F. The greater the temperature difference between the primary and secondary coolant flowing through the heat exchangers of the steam generating vessel 300, the greater the thermal siphon effect and circulation rate of the primary coolant through the primary coolant flow loop. The vertical orientation of the steam generating vessel 300 and reactor vessel 200 further beneficially contribute to the thermal gravity primary coolant flow based generally on the principle that hot water rises and cold water falls. With reference to FIG. 6, the temperature differential between the primary coolant leaving the steam generating vessel 300 (at the preheater 320) and liquid secondary coolant (feedwater) entering the steam generating vessel (at the preheater) preferably is sufficient to induce and boost the convective thermally-driven gravity circulation of the primary coolant through the steam generating vessel and reactor vessel 200. In one embodiment, the temperature difference between the primary coolant leaving the steam generating vessel and secondary coolant entering the steam generating vessel is at least 175 degrees F. which boosts the thermal-siphon gravity flow effect. An acceptable range of the liquid secondary coolant feedwater leaving feedwater heater 540 may be between about and including 170-210 degrees F. which flows directly to the steam generating vessel 300. An acceptable corresponding range of primary coolant leaving the preheater section 320 of the steam generating vessel 300 may be between about and including 365-405 degrees F. The nuclear steam generating system and Rankine cycle operation according to the present disclosure will now be summarized. On the secondary coolant side, the method includes converting the secondary coolant from liquid phase to superheated steam using the heat exchangers in the steam generating vessel in the manner as already described. To summarize, liquid secondary coolant flows and enters the preheating section 320 from the external portion of the secondary coolant flow loop 500 directly from the feedwater heater 540 without any intervening heating. The secondary coolant has a first inlet temperature. There are no other heaters in the external loop which perform heating of the liquid secondary coolant between the steam turbine and the inlet to the steam generating vessel (noting that the condenser 420 further cools the secondary coolant). The liquid secondary coolant flows upwards through the steam generating vessel 300 first through the steam generator section 330 where it is converted to steam, then finally through the superheater section 350 where the steam becomes superheated in state. The superheated steam exits the steam generating vessel 300 at a higher second outlet temperature than the first inlet temperature and re-enters the external portion of the secondary coolant flow loop. On the primary coolant side, the primary coolant from the reactor vessel initially enters the steam generating vessel at a first temperature and flows upwards to the pressurizer 380 and enters the tube-side of the superheater section 350. The primary coolant flows downwards in order through the superheater section 350, steam generator section 330, and preheater section 320 where the primary coolant exits the steam generating vessel 300 and returns to the reactor vessel 200. The primary coolant, having lost and transferred heat to the secondary coolant to convert the secondary coolant from liquid to superheated steam, exits the steam generating vessel 300 at a second lower outlet temperature than the inlet temperature. The secondary coolant is cooled in the external portion of the secondary coolant flow loop 500. The secondary coolant leaving the steam generating vessel 300 in the form of superheated steam flows into the low pressure steam turbine 900 at a pressure less than 400 psia in one embodiment. The corresponding steam temperature may be may be at least 575 degrees F. in one embodiment, as already described above. The steam flows through the turbine 900, the condenser 520, feedwater pump 530, feedwater heater 540, and directly back to the inlet of the steam generating vessel 300 and preheater section 320. The secondary coolant has lost a significant portion of its heat energy (and thus temperature) to drive the turbine which is converted into electrical energy. Unless otherwise specified, the components described herein may generally be formed of a suitable material appropriate for the intended application and service conditions. All conduits and piping are generally formed from nuclear industry standard piping. Components exposed to a corrosive or wetted environment may be made of a corrosion resistant metal (e.g. stainless steel, galvanized steel, aluminum, etc.) or coated for corrosion protection. 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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047568755 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT The filtering apparatus of this invention has a construction as shown in FIG. 2. As shown, the apparatus comprises a cylindrical casing 5 containing cylindrical partition walls 7 and a head plate 12 from which a plurality of U-shaped bundles 3 of hollow fibers having a sponge-like mesh structure are suspended. Cap 20 is located on the opposite side of head plate 12 from casing 5 and defines collection chamber 22 which is in communication with the hollow interior of the fibers in bundles 3. Water containing radioactive impurities (CRUD) is admitted into the casing 5 through a pipe 8 and filtered while passing through numerous pores extending in the transverse direction of each hollow fiber. Cleaned water passes from the hollow interior of the fibers into collection chamber 22, from which it is discharged to the outside through pipe 9. The casing 5 is supported by a plurality of pedestals 13. To effect backwashing, pressure is applied to the interior of the casing 5, by pressurized air, for example. Then clean water is forced to pass through the pores to remove small solid particles trapped therein or large solid particles closing the outside openings of the pores. The solid particles thus removed are discharged to the outside of the casing 5 through a drain pipe 14 at the bottom thereof. At the time of backwashing, it is advantageous to blow air bubbles into the water in the casing 5 through air nozzles 6 provided for an air inlet pipe 4 near the lower ends of respective U-shaped bundles to vibrate the hollow fibers in various directions as shown in FIG. 4 to facilitate removal of the solid particles. Air collected beneath the head plate 12 is discharged to the outside through a discharge pipe 11. According to this invention, the hollow fibers are made of flexible polyethylene. The following Table shows comparison of various mechanical characteristics of hollow fibers made of flexible polyethylene and polyvinyl alcohol. TABLE ______________________________________ Sample Polyvinyl Item Polyethylene alcohol ______________________________________ pull strength 336 253 (g/fiber) elongation 45.2 7.6 (%) knot strength 319 impossible (g/fiber) to measure elongation 48.3 because PVA (%) fiber is hang strength 568 brittle (g/fiber) elongation 32.0 (%) ______________________________________ Remarks: As diagrammatically shown in the Table, in the pull test one end of a fiber was secured and the other end is pulled; in the knot test a knot is formed at an intermediate point of the fiber; and in the hang test one end of one U-shaped fiber is fixed, the other U-shaped fiber is hung from the one U-shaped fiber. The tested flexible polyethylene had a pore size sufficient to pass solid particles having a molecular weight of about 300,000. As can be noted from this Table the flexible polyethylene hollow fibers utilized in this invention have higher mechanical strength than conventional polyvinyl alcohol hollow fibers. It should be particularly noted that the polyethylene hollow fibers utilized in this invention are pliable or flexible, so that there are such advantages that they would not be broken by external force so that they can be handled readily and that as shown in FIG. 4 they can bend in various directions during backwashing due to a turbulent flow caused by air bubbles thereby increasing removal of trapped solid particles. In contrast, since polyvinyl alcohol hollow fibers are solid and brittle they can not manifest these advantages. As the filtering operation proceeds, the pressure difference between the inlet pipe 8 and the discharge pipe 9 increases as a result of trapping or clogging of the solid particles in the pores, so that when the pressure difference reaches a predetermined value, backwashing is carried out in a manner described above. FIGS. 5 and 6 are graphs showing the relation between the number of backwashings and the variation in the pressure difference when hollow fibers having pore diameters of 1 micron and 0.1 micron respectively are used. Comparison of FIGS. 5 and 6 shows that, in the case of pore diameter of 1 micron, the rise in the pressure difference after backwashing is large (about 3.0 Kg/cm.sup.2 or more), whereas in the case of pore diameter of 0.1 micron, the rise in the pressure difference is not so large (only about 1 Kg/cm.sup.2). This means that with hollow fibers having a pore diameter of 1 micron, clogging is liable to occur but with hollow fibers having a pore diameter of 0.1 micron, the tendency of clogging is small with the result that the frequency of renewal of the hollow fibers can be reduced. As a consequence, hollow fibers having a pore diameter of about 0.1 micron are suitable for filtering water containing radioactive solid particles. |
description | The present invention is a system and method for generating electricity by combining a fuel core and a drive regulation and containment system, the fuel core having a plurality of radioactive isotopes disposed between a plurality of crystalline lattices, and the drive regulation and containment system having a plurality of electromagnets that concentrate charged particles generated in the fuel core from the plurality of radioactive isotopes, and an electric field generated by an electron flow initiation system for driving the charged particles through the fuel core to create a current flow. These and other embodiments of the present invention will become apparent to those skilled in the art from a consideration of the following detailed description taken in combination with the accompanying drawings. Reference will now be made to the drawings in which the various embodiments of the present invention will be given numerical designations and in which the embodiments will be discussed so as to enable one skilled in the art to make and use the invention. It is to be understood that the following description illustrates embodiments of the present invention, and should not be viewed as narrowing the claims which follow. FIG. 1 is a cut-away profile view of a first embodiment of a system and method for generating electricity from the radioactive decay of radioactive isotopes. The energy generating system 30 may be comprised of a fuel core 32, a drive regulation and containment system 34 and a housing 36. Coupled to the fuel core 32 is an energy flow or electron flow initiation system 38. The electron flow initiation system 38 may be used to start the flow of electrons from the fuel core 32 and provide power to an electrical load or system 50. The electron flow initiation system 38 may also affect a rate of the flow of electrons from the fuel core 32. The direction of travel of electrons into the fuel core 32 is indicated by the arrow 52, and out of the fuel core by the arrow 54. The electron flow initiation system 38 may be any power source that provides an electric field through the fuel core 32. It should be understood that the power required to initiate electron flow through the fuel core 32 is very small compared to the electron flow from the fuel core. The components shown in figure are for illustration purposes only and should not be considered to be limiting the structure of the energy generating system 30. The energy generating system 30 may be modified without departing from the inventive aspects of the present invention. The exact size of the components, the relative size of the components, the spacing between the components and the precise shape of the components may be modified from what is shown and still fall within the inventive aspects of the disclosure. FIG. 2 is a cut-away profile end view of the energy generating system 30 of FIG. 1. The end view shows that components of the drive regulation and containment system 34 are disposed on all sides of the fuel core 32. An explanation of the components of the energy generating system 30 now follows. FIG. 3 is a top view of a first embodiment of a component of the fuel core 32. In this first embodiment, the fuel core 32 may be comprised of a crystalline lattice and a plurality of radioactive isotopes 44. FIG. 3 is a top view of a portion of one possible shape of a crystalline lattice 40. The crystalline lattice 40 may have properties that enable it to absorb charged particles from the energy released by the radioactive decay from the plurality of radioactive isotopes 44 in the fuel core 32. The crystalline lattice 40 may also function as a frame for holding the plurality of radioactive isotopes 44. The crystalline lattice 40 shown in FIG. 3 may be a crystalline lattice 40 formed from Graphene, with edges of the six sided lattice structures being approximately 0.35 nm. However, this size limitation should not be considered as limiting. The lattice structure of other materials may be different, but may be within an order of magnitude of these dimensions. FIG. 4 is a cut-away profile view of a stack of the crystalline lattices 40. The stack of the crystalline lattices 40 creates locations 42 in which the plurality of radioactive isotopes 44 may be disposed between the layers of the stack. Thus, the size and shape of the locations 42 may be increased or decreased as necessary in order to properly hold the radioactive isotopes that are selected for use on the fuel core 32. It should be understood that the plurality of radioactive isotopes 44 may or may not fill all of the locations 42 in the stack of the crystalline lattices 40. The stack of the crystalline lattices 40 may be formed into a plurality of layers for holding the plurality of radioactive isotopes 44, but also for absorbing charged particles that are emitted from the plurality of radioactive isotopes. The crystalline lattice 40 may undergo absorption of alpha and beta radiation and free electrons given off by the decay of the plurality of radioactive isotopes 44. The charged particles may be positively or negatively charged. For example, the charged particles may be positrons or electrons (referred to hereinafter only as “electrons”). The different layers of the crystalline lattices 40 may not touch each other, but may instead be kept apart by the source of charged particles (the plurality of radioactive isotopes 44). The shape of the fuel core 32 may be any two-dimensional or three-dimensional shape that provides the desired function. Some shapes may make containment more convenient or increase current flow for a given volume of space. For example, the cross-section of the fuel core 32 may be triangular, rectangular or circular, or it may not be limited to any of these shapes. The crystalline lattices 40 may have Dirac properties. For example, the crystalline lattices 40 may take the form of Dirac cones. Dirac cones are features in the band structure of a two-dimensional material where the conduction and valence bands meet in a single point in the Fermi level. The bands approach this point in a linear way, meaning that the effective kinetic energies of the conduction electrons (and holes) are directly proportional to their momenta. This unusual relationship is normally only seen for photons, which are massless, because the energies of the electrons and other particles of matter at non-relativistic velocities usually depend on the square of their momenta. The result is that the electrons in the Dirac cones may behave as though they are relativistic particles with no rest mass, traveling through the material at extremely high speeds. The crystalline lattices 40 with Dirac cone properties may absorb the free-flowing electrons that are available from the active particles source (the plurality of radioactive isotopes 44). The crystalline lattices 40 may have near lossless properties (lossless conductivity of particles) and may possess favorable electrical and thermal conductive properties. The active particle source may supply an excess amount of electrons while the crystalline lattices 40 may provide a structure to harvest the available free electrons. A source of charged or active particles may be associated with any type of radioactive isotopes. For example, Strontium-90 has both high beta radiation and elevated electron activity, and may be used as the active particle source. However, Strontium-90 is only an example and should not be considered as limiting of the different radioactive isotopes that may be used in the fuel core 32 of the present invention. Another feature of the crystalline lattices 40 of the fuel core 32 is that they may be scalable. A scalable structure for the crystalline lattices 40 may enable the amount of current flow and thus the total amount of electricity to be generated from the fuel core 32 to substantially vary. Thus the number of layers in the stack of the crystalline lattices 40 may be increased or decreased, or the very structure such as the size of the locations 42 in the crystalline lattices 40 may be increased or decreased. Examples of materials having Dirac crystalline lattice structures include, but should not be considered as limited to, Graphene, Bismuth-Antimony and Boron. However, any material may be used for the crystalline lattices 40 that can provide the desired features of the fuel core 32. The housing 36 of the energy generating system 30 may help to contain the decay reaction of the plurality of radioactive isotopes 44, it may contain a pressure within the fuel core 32, or it may do both functions. The electron flow initiation system 38 may pass through the housing 36 in order to gain access to the fuel core 32. The drive regulation and containment system 34 of the energy generating system 30 enables the charged particles in the stack of crystalline lattices 40 to be moved through the lattice structure, thereby creating current flow and the generation of electricity. The drive regulation and containment system 34 may be comprised of a plurality of magnets as shown in FIGS. 1 and 2. The plurality of magnets 34 may be used to control a direction of current flow through the fuel core 32, as well as a rate at which the current flows through the fuel core. The plurality of magnets 34 may be permanent which may make it difficult to control current flow. Accordingly, a plurality of electromagnets 34 may enable control of current flow. It may also be possible to use a combination of permanent and electromagnets as the drive regulation and containment system 34. The plurality of magnets, permanent and/or electromagnetic, may be configured around a perimeter of the fuel core 32. With the plurality of magnets 34 on all sides of the fuel core 32, the charged particles may be compressed into a center of the fuel core. The charged particles may form a shape that is determined by the magnetic field lines of the plurality of magnets 34. Thus if the plurality of magnets 34 are in a linear arrangement along a length of the fuel core 32 as shown in FIG. 1, the charged particles may be formed into a linear arrangement. This arrangement of charged particles may be referred to herein as a tunnel. The tunnel may or may not be linear in shape. An example of a specific geometry for the plurality of magnets 34 is a quadrupole design as known to those skilled in the art. One example of a quadrupole is shown in FIGS. 1 and 2. However, the plurality of magnets 34 is not limited to this structure as will be shown. FIG. 5 is a more detailed perspective view of a plurality of electromagnets 34 that are disposed around the fuel core 32. This figure is an illustration of a linear fuel core 32 and drive regulation and containment system 34. Just as an example of the size of the structure being shown, the length of the fuel core 32 and the plurality of electromagnets 34 around the fuel core is approximately 1.5 feet. However, these dimensions shown are for illustration purposes only, and should not be considered as limiting. The dimensions of the energy generating system 30 may vary greatly because of the scalability of the system. Thus, the length of the energy generating system 30 may be less than one inch or may be greater than 10 feet without departing from the teachings of the present invention. FIG. 6 is a detailed and close-up perspective view of a single one of the plurality of electromagnets 34 that may be used to create the drive regulation and containment system shown in FIG. 5. This figure is for illustration purposes only and should not be considered as limiting of the shape of the electromagnets. FIG. 7 is a perspective view of another embodiment of a fuel core 32 and drive regulation and containment system 34 of the energy generating system 30. The fuel core 32 and the plurality of electromagnets 34 may be cylindrical as shown. The fuel core 32 is shown as a cylindrical tube disposed between an outer layer of electromagnets 34 and an inner layer of electromagnets. In all embodiments, it is presumed that an end cap 56 may be disposed at both ends of the fuel core 32 and the plurality of electromagnets 34. The end cap 56 may be part of the containment system 36 or it may be a separate component. FIG. 8 is a perspective view that illustrates that one method for creating the stack of crystalline lattices 40 may be to create a single sheet of crystalline lattices 40 that is rolled into a cylindrical shape as shown. The efficiency of a cylindrical shape may or may not increase energy production of the energy generating system 30. FIG. 9 is a perspective view of another embodiment of the present invention, where sheets 46 that function as the crystalline lattices 40 are twisted together in a helical design of the fuel core 32. It should be understood that there are several layers of the crystalline lattices 40 that are twisted to form a multi-layer helical design of the fuel core 32. One aspect of the present invention that may be useful to understand further is the path of the charged particles through the fuel core 32. The path followed by the charged particles through the fuel core 32 may or may not be linear. When the fuel core 32 is linear, the path is likely to be linear, but may not always be so. The shape of the path of the charged particles may conform to the shape of the fuel core 32, the plurality of crystalline lattices 40, or it may be different. The shape of the path may be determined by the plurality of electromagnets 34. For example, in a cylindrical fuel core 32, the path of the charged particles may be helical. The present invention may function as long as there is a path for the charged particles through the fuel core between the ends of the electron flow initiation system 38. While the plurality of magnets 34 form the tunnel for the charged particles, and the tunnel may flow through the stack of crystalline lattices 40 of the fuel core 32, an electric field may be used to direct the flow of the charged particles through the tunnel. The direction of the flow of the charged particles may be changed by changing the electric field. The electric field from the electron flow initiation system 38 that may be used to start and also to drive the current through the fuel core 32 may also provide excitation energy so that electrons in the plurality of radioactive isotopes 44 may reach the outer energy shell as the electromagnetic field may excite the decay rate of the plurality of radioactive isotopes. Thus, a magnetic field from the plurality of electromagnets 34 may concentrate the charged particles, while an electric field from the electron flow initiation system 38 may provide the direction of flow as the charged particles are harvested from the fuel core 32. The decay rate of the plurality of radioactive isotopes 44 may also be influenced by pressure within the fuel core 32. Accordingly, another aspect of the present invention is to create the fuel core 32 such that the pressure within may be increased or decreased in order to control the decay rate. Therefore, the fuel core 32 may be sealed in order to control pressure within. The concentration of the charged particles in the fuel core 32 may be affected by changing the characteristics of the plurality of electromagnets 34 that are also providing containment. Thus, the concentration of the charged particles may be influenced by the strength, position and size of the plurality of electromagnets 34, the strength of the electron flow initiation system 38, the volume of the fuel core 32, and the magnitude of the charged particles coming from the plurality of radioactive isotopes 44. It should be understood that the electron flow initiation system 38 is attached at a first end 52 of the fuel core 32 to drive the electrons from the fuel core at a second end 54 to any electric load or system 50, including an electrical circuit, an electrical motor or an energy storage system such as a battery or capacitor. Although only a few example embodiments have been described in detail above, those skilled in the art will readily appreciate that many modifications are possible in the example embodiments without materially departing from this invention. Accordingly, all such modifications are intended to be included within the scope of this disclosure as defined in the following claims. It is the express intention of the applicant not to invoke 35 U.S.C. §112, paragraph 6 for any limitations of any of the claims herein, except for those in which the claim expressly uses the words ‘means for’ together with an associated function. |
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abstract | Illustrative embodiments provide nuclear fission igniters for nuclear fission reactors and methods for their operation. Illustrative embodiments and aspects include, without limitation, a nuclear fission igniter configured to ignite a nuclear fission deflagration wave in nuclear fission fuel material, a nuclear fission deflagration wave reactor with a nuclear fission igniter, a method of igniting a nuclear fission deflagration wave, and the like. |
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039986918 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS According to the present invention, at first, about 10g of telluric acid is held at a temperature in the range of 500.degree. - 550.degree. C for about 2 hours in an electric furnace in order to prepare a tellurium oxide intermediate. The tellurium oxide intermediate thus obtained is exposed to a neutron flux of a suitable intensity for a suitable period (Irradiation in an nuclear reactor may be convenient from the viewpoint of its supplying a neutron flux with a high density compared with other sources of the flux), then it is transferred to a quartz tube. One end of the quartz tube is connected to a gas absorbing bottle which contains sodium hydroxide aqueous solution with a suitable concentration, the other end being connected to a pump which supplies air to the tube. While supplying air by means of the pump to the quartz tube, the surface of the tellurium oxide intermediate and the gas absorbing bottle in order, the tellurium oxide intermediate is heated to above its decomposition temperature (about 560.degree. C), preferably to about 600.degree. C, and held at that temperature for about two hours in an electric furnace or the like. The final product, I-131, is recovered from the gas absorbing bottle in the form of sodium iodide aqueous solution. In this process, about 80% of the formed I-131 in the irradiated tellurium oxide intermediate is recovered after the first 1 hour, and about 97% is recovered after the first 2 hours. When the temperature is held at about 600.degree. C, no contamination of the final product by volatilized tellurium is detected. EXAMPLE About 10 grams of telluric acid, a commercially available chemical, was placed in an electric furnace and was held at 500.degree. C for 2 hours. The decrease in the material by heating was confirmed to be 2.9 grams per 10 grams of said material. Tellurium oxide intermediate, which is used for the preparation of I-131 according to this invention, was thus obtained. One (1) gram of the tellurium oxide intermediate was enclosed in a quartz tube 0.8 cm in diameter and 5 cm in length, and was placed in a cylindrical aluminum container 2.5 cm in diameter and 11.4 cm in length, for irradiation. The enclosed material was irradiated by a neutron flux of a density of 3 .times. 10.sup.13 n/cm.sup.2.sup.. sec for 10 days in a heavy water-cooled CP-5 type nuclear reactor which was operated with the power of 10 MW. After irradiation, the irradiated material was taken out and was inserted in a quartz tube 2.0 cm in diameter, which was positioned at the center of a cylindrical electric furnace. One end of the quartz tube was connected to a gas absorbing bottle, which contains 20% of 0.5% sodium hydroxide aqueous solution. The other end of the quartz tube was connected to an air pump, from which air was blown through the tellurium oxide intermediate to the absorbing bottle at the flow rate of 50-100 cc/min. The material was held at 600.degree. C for 2 hours. By this process, 97% of the I-131 in the irradiated material was obtained as sodium iodide aqueous solution in the absorbing bottle. The thus obtained I-131 corresponded to the radioactivity of 50 m Ci at the end of the reactor irradiation. No tellurium was detected in the final product. The "tellurium oxide intermediate", herein used, begins to decompose at 560.degree. C and simultaneously releases I-131 as mentioned precisely latter. This material is more stable, when compared with tellurium trioxide that begins to decompose at 400.degree. C, under the irradiation in severe conditions, namely, with a neutron flux of high density and for a long time. This indicates that said material yields a large amount of I-131 and is useful as a starting material for producing I-131. The temperature at which the material begins to release I-131 is about 560.degree. C (this is higher by about 100.degree. C than the temperature required for tellurium trioxide, namely, 450.degree. C). It is not difficult to work out at such a temperature. Actually, even though it is heated at about 600.degree. - 650.degree. C in order to recover I-131 efficiently, tellurium does not volatilize as does tellurium dioxide when it is used, and pure I-131 can be obtained. The above preferable embodiment and the example are indicated for the precise explanation of this invention and not to restrict same. It should be certainly understood by one skilled in the art that this invention covers such modifications and variations as come within the spirit and the scope of the appended claims. For further understanding of this invention, the following main embodiments and their theoretical basis are indicated. 1. The material, tellurium oxide intermediate, is obtained by heating commercially available telluric acid at about 500.degree. C, passing through tellurium trioxide in the course of the decomposition process of telluric acid. Therefore, commercially available tellurium trioxide may be used as a starting material for preparing the tellurium oxide intermediate. 2. The conditions of irradiation by a nuclear reactor can be varied within the scope, as the crystal structure of the material is not changed. 3. I-131, formed in the irradiated tellurium oxide intermediate, is sent to a gas absorbing bottle by means of an air stream. However, the gas is not limited to air and other suitable gases can be used for this purpose, since it is one of the means to send I-131 to a gas absorbing bottle. 4. The sodium hydroxide aqueous solution contained in the gas absorbing bottle is used for collecting I-131 in the form of alkali metal iodide, and, therefore, an aqueous solution of carbonate each of sodium and potassium, thiocarbonate each of sodium and potassium, which have the same effect as said sodium hydroxide does, or a mixture thereof may be also used for this purpose. Also such solution may contain a sulfite or thiosulfide which stabilizes the collected I-131 chemically. 5. Even though distillation of I-131 (namely, isolation of I-131 in the irradiated material from same) begins at about 560.degree. C, for practical purpose it is preferable to heat it at about 600.degree. C to 650.degree. C in order to distil I-131 efficiently. It is useless to heat above 650.degree. C since, by doing so, tellurium will begin to volatilize and no substantial advantage can be obtained since the temperature distribution in an electric furnace is not uniform at such a high temperature. Tellurium trioxide, TeO.sub.3, (i.e. commercially available tellurium trioxide or one which is obtained by heating commercially available telluric acid at about 300.degree. C) begins to decompose above 400.degree. C and converts itself into tellurium dioxide, TeO.sub.2, above 560.degree. C, liberating oxygen. The inventors have found that a tellurium oxide (i.e. a tellurium oxide intermediate used in this invention), which is stable for a temperature ranging from about 500.degree. C to about 560.degree. C, can be obtained by carefully heating said tellurium trioxide and maintaining same at about 500.degree. C for a suitable period of time, preferably for 1 to 2 hours. The composition of the oxide can be expressed at TeOn, wherein n is between about 2.22, i.e. 2TeO.sub.3.7TeO.sub.2, and about 2.18, i.e. 2TeO.sub.3 .9TeO.sub.2. Said chemical composition is intermediate between those of TeO.sub.3 and TeO.sub.2, wherein a part of the crystal structure of TeO.sub.3 is converted to that of TeO.sub.2 by the release of oxygen accompanying the decomposition. It is confirmed by X-ray diffraction analysis that the crystal structure of the tellurium oxide intermediate used in this invention differs from that of TeO.sub.2 or TeO.sub.3. Accordingly, it indicates that said tellurium oxide intermediate is not a mere mixture of TeO.sub.3 and TeO.sub.2. Even though TeO.sub.3 is carefully heated, it is difficult in practice to obtain a compound, wherein the composition of which is TeO.sub.3.nTeO.sub.2 and n is constant. Usually n varies between 3.5 and 4.5 as mentioned above. This seems to be caused by the lack of uniformity of temperature in the furnace when TeO.sub.3 is heated in an electric furnace. In any case, the compound begins to decompose releasing oxygen at about 560.degree. C and converts to TeO.sub.2. If such compound, i.e. the tellurium oxide intermediate, is irradiated by a neutron flux in a nuclear reactor and then is heated, it actually begins to release I-131 near the temperature of 500.degree. C. This results from the diffusion of I-131 atoms enclosed in the crystal lattice of the irradiated tellurium oxide intermediate. The rate of the release is slow and, therefore, it is not practically useful. When the tellurium oxide intermediate is heated gradually above a temperature of about 500.degree. C, the rate of the release of I-131 brought about by the diffusion increases with the increase in temperature, and when the temperature reaches about 560.degree. C, the release of I-131 suddenly becomes rapid. At this temperature, the tellurium oxide intermediate converts to tellurium dioxide as mentioned above, and simultaneously the solid geometrical rearrangement of the component atoms (crystal structure transition) occurs within the crystal. It can be explained as follows: as the result of said reaction brought about by the irradiation of the tellurium oxide intermediate with a neutron flux, I-131 atoms are produced within the irradiated material, and, although the I-131 at ms included in the crystal lattice of the material are observed to be released gradually by diffusion in solid phase above the temperature of about 500.degree. C, said atoms are suddenly released rapidly from the crystal lattice at the temperature of 560.degree. C, accompanying the thermal decomposition of the irradiated material and the following crystal structure transition at that temperature. In a practical process, the tellurium oxide intermediate is preferably heated at a temperature in the range of from about 600.degree. C to about 650.degree. C in order to promote the decomposition of said material, and, consequently, to obtain I-131 effectively. However, heating above 650.degree. C should be avoided since it may bring about the volatilization of tellurium. |
abstract | Continuous wave laser apparatus with enhanced processing efficiency is provided as well as a method of manufacturing a semiconductor device using the laser apparatus. The laser apparatus has: a laser oscillator; a unit for rotating a process object; a unit for moving the center of the rotation along a straight line; and an optical system for processing laser light that is outputted from the laser oscillator to irradiate with the laser light a certain region within the moving range of the process object. The laser apparatus is characterized in that the certain region is on a line extended from the straight line and that the position at which the certain region overlaps the process object is moved by rotating the process object while moving the center of the rotation along the straight line. |
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047986999 | summary | CROSS REFERENCE TO RELATED APPLICATION Reference is hereby made to the following co-pending U.S. patent application dealing with subject matter related to the present invention: "Control Rod End Plug with Stabilizing Configuration" by John F. Neilson et al, U.S. Ser. No. 634,729, filed July 26, 1984. BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactors having rods, such as of the control or water displacer type, reciprocable into and out of a reactor core and, more particularly, is concerned with a wear sleeve composed of the same material as the guide thimble within which the rod reciprocably moves and attached to an end plug on the rod composed of material dissimilar to the guide thimble material such that the wear sleeve provides the contact interface between the rod end plug and the guide thimble. 2. Description of the Prior Art In a typical nuclear reactor, the reactor core is composed of a plurality of elongated fuel assemblies each of which contains a plurality of elongated fuel elements or rods. A liquid coolant is pumped upwardly through the core in order to extract heat generated in the core for the production of useful work. The heat output of the core is usually regulated by the movement of control rods containing neutron absorbing material such as B.sub.4 C, or by movement of water displacer rods such as the ones described in U.S. Pat. No. 4,432,934. In reactors of the pressurized-water type, each fuel assembly typically includes a plurality of cylindrical guide tubes or thimbles through which the cylindrical control rods or water displacer rods are reciprocably moved. Some of the coolant flow is usually diverted into the lower end of the guide thimble in order to cool the control rod. The control rod ordinarily generates heat in the nuclear transformation associated with its neutron absorbing function. During power operation of the reactor, most of the regulating control rods are maintained substantially withdrawn from the reactor core and thus disposed in withdrawn positions in which the lower end plug tips of the control rods are within the upper ends of the guide thimbles. While in such withdrawn positions, the control rods may experience significant vibration induced by coolant water flow within the guide thimbles which results in oscillatory contact of the rod end plug tips against the internal wall surfaces of the guide thimbles and wear on these surfaces. Continuous wear of the guide thimble walls can lead to perforation of the thimbles and significant weakening of the fuel assembly structure. Thus, there has arisen the need to significantly mitigate the affects of the control rod vibrations so as to bring guide thimble wear under control. Two approaches to solving this problem are disclosed in U.S. Pat. Nos. to Schukei et al (U.S. Pat. No. 4,292,132) and Verdone (U.S. Pat. No. 4,311,560). Both of these approaches have as a common objective the elimination of wear on the guide thimble wall by preventing vibratory contact of the control rod against its adjacent guide thimble wall. In the Verdone approach, a spring device is added to the lower end of the control rod which provides a uniform, resilient interference fit against the guide thimble wall and thereby prevents the rod tip from impacting the guide thimble wall. In the Schukel et al approach, the control rod has a hydraulic bearing formed at its lower tip which produces forces which counteract forces tending to drive the control rod tip against the guide thimble wall. In such manner, contact of the control rod against the guide thimble wall and resultant wear thereon are substantially avoided. While the approaches taken in these two patents operate reasonably well and achieve their objectives under the range of operating conditions for which they were designed, a need exists for an alternative approach to the wear problem which is simplier and less costly in its design and construction and is more reliable in its performance over the long term. SUMMARY OF THE INVENTION The present invention provides a wear sleeve on the lower end plug of the rod designed to satisfy the aforementioned needs. Unlike the prior art approaches which prevent vibratory contact with the guide thimble wall by interposing a device which either maintains continuous contact with the wall or prevents any contact at all from occurring, the present invention allows vibratory contact with the guide thimble wall, but provides the insert sleeve at the contact interface. Since the rate of wear is influenced to a large degree by the materials at the contacting interface, the sleeve in being composed of the same material as the guide thimble greatly reduces wear potential at the contact interface over that experienced heretofore between the dissimilar materials of the thimble and the rod lower end plug. The wear sleeve is simply and reliably attached to the rod end plug by means of a single circumferential countersink or protuberance swaged or mechanically rolled into the wear sleeve so as to extend into an annular groove circumferentially formed in the rod end plug. Accordingly, the present invention sets forth in a nuclear reactor including a plurality of upstanding guide thimbles, a plurality of control rods received in the guide thimbles and means supporting the control rods for movement relative to the thimbles between inserted and withdrawn positions, a wear sleeve disposed on an end plug attached to an end of each tubular cladding member of the control rod for providing a contact interface between the control rod and its respective guide thimble. On the one hand, the end plug and cladding member of the control rod are formed of the same material, such as zircaloy, while, on the other hand, the wear sleeve and guide thimble are formed of the same material, such as stainless steel. The end plug is rigidly attached and sealed to the end of the cladding member by a girth weld. Preferably, the wear sleeve is inserted about the end plug to a position where an inner end of the sleeve is spaced a short distance from the girth weld. The wear sleeve has generally the same outside diameter as the cladding member. Attachment of the sleeve is accomplished by an interlock connection in the form of a single circumferential countersink swaged mechanically rolled at a location intermediately between the ends of the sleeve into a circumferential groove formed near the inner end of the end plug. A single swaged countersink or protuberance is desirable in order to eliminate axial thermal mismatch between the dissimilar materials of the sleeve and the end plug. In the alternative, the sleeve is mechanically rolled so as to capture the sleeve on the end plug and provide a diametric gap therebetween such that some relative diametric movement is available in creating a dampening effect. An interlock connection is used, rather than a weld, because dissimilar metals cannot be welded together. These and other advantages and attainments of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
039716996 | summary | The present invention relates to neutronic reactors with solid moderators, and to methods of operating such reactors which minimize the deleterious effect of high energy neutron bombardment of the moderator over prolonged periods of time. It has been found that prolonged operation of a neutronic reactor with a solid moderator will cause changes in the physical properties of the material of the moderator, particularly when the reactor is operated at a relatively high power level. Among the changes which occur in the moderator, expansion presents the most immediate problems, since distortion of the shape of the reactor limits the life of the reactor. Extruded moderator materials expand normal to the axis of the extrusion of the material. For this reason, spaces have been provided adjacent to blocks of moderator material on the sides parallel to the axis of extrusion, as described in the copending application of John T. Carleton, Ser. No. 157,287, filed Apr. 21, 1950 now U.S. Pat. No. 3,200,046 dated Aug. 10, 1965. The changes in the physical properties of solid materials as a result of prolonged high energy neutron bombardment are believed to be due to distortion of the crystalline lattice of such materials. When a high energy neutron strikes one of the atoms of a crystal, the bonds which tie the atoms of the crystal together are placed under severe strain, and under certain conditions one of the atoms will be displaced from its position in the lattice structure. This may merely cause rearrangement of the atoms in the lattice structure of the crystal, or it may result in the bombarded atom entirely leaving the lattice structure and taking up a permanent interstitial position. This latter process is believed to be the one which occurs when crystalline materials are expanded by neutron bombardment. It has been found, that the expansion of crystalline materials under neutron bombardment may be minimized if the crystalline materials are maintained at a sufficiently high temperature in order to facilitate "self-healing." Self-healing is the process which occurs when the temperature at which the bombarded crystalline material is maintained sufficiently high to facilitate atoms in resuming the normal crystalline lattice structure. This method of curing the deformations in a crystalline substance was first discovered by Eugene P. Wigner, and described in the copending patent application, Ser. No. 605,958, filed July 19, 1945 now abandoned. A neutronic reactor constructed with a solid crystalline moderator which minimizes the moderator expansion by means of self-healing is disclosed in the copending patent application of Alfred A. Johnson and John T. Carleton, Ser. No. 253,908, filed Oct. 30, 1951. The present invention retards the expansion of the solid crystalline materials under neutron bombardment by maintaining the crystalline material in a fluid atmosphere of a fluid selected to have a thermal conductivity which reduces the heat transfer from the crystalline moderator material to whatever coolant medium is present, thereby increasing the temperature of the crystalline material. As a result, the rate of thermal self-healing will be increased. It is thus an object of the present invention to provide a neutronic reactor with a solid crystalline moderator in which expansion of the moderator is decreased or inhibited. It is also an object of the present invention to provide a method of operating neutronic reactors with crystalline moderators which will decrease or minimize the expansion of the moderator. |
description | This application claims the benefit of Korean Patent Application No. 2009-0001116, filed on Jan. 7, 2009, entitled “PERFORATED PLATE SUPPORT FOR DUAL-COOLED SEGMENTED FUEL ROD,” which is hereby incorporated by reference in its entirety into this application. 1. Field of the Invention The present invention relates, in general, to a support for a dual-cooled segmented fuel rod and, more particularly, to a perforated plate support for a dual-cooled segmented fuel rod, capable of stably supporting the fuel rod to the end of its cycle even if an gap between the fuel rods becomes narrow due to application of a dual-cooled fuel rod, and of reducing vibration induced by flows of the inside and outside of the dual-cooled fuel rod for obtaining high burnup and power. 2. Description of the Related Art A nuclear fuel assembly is charged in the core of a pressurized water reactor. This nuclear fuel assembly is composed of a plurality of fuel rods, in each of which a cylindrical uranium sintered compact (or a cylindrical uranium pellet) is inserted. The fuel rods can be divided into two types, cylindrical and annular, according to shape. The annular fuel rods are called dual-cooled fuel rods. In comparison with the pellet of the cylindrical fuel rod, the pellet of the annular fuel rod has a low internal temperature due to a thinner thickness and a wider heat transfer area, and thus a relatively higher safety margin. FIG. 1 is a schematic front view illustrating a conventional cylindrical nuclear fuel assembly. Referring to FIG. 1, the nuclear fuel assembly 100 includes fuel rods 101, spacer grids 105, guide thimbles 103, an upper end fitting 107 and a lower end fitting 106. Each fuel rod 101 has a structure in which a uranium sintered compact or a uranium pellet (not shown) generating high-temperature heat through nuclear fission is enclosed by a zirconium alloy cladding tube. Each fuel rod 101 has upper and lower end plugs 108 and 109 coupled to lower and upper portions thereof so as to prevent inert gas filled between the cladding tubes thereof from leaking out. Meanwhile, the structure of the fuel rod 101 has a length considerably long compared to the diameter thereof When this structure having a great slenderness ratio is subjected to coolant flow, the fuel rod 101 causes flow-induced vibrations due to the flow of the coolant. Thus, in order to reduce these flow-induced vibrations, the structure called a spacer grid 105 is installed in a predetermined section with respect to the entire length of the fuel rods 101 so as to support the fuel rods 101, thereby preventing the fuel rods 101 from being vibrated by the flow of the coolant. However, in the case of the dual-cooled fuel rod designed to charge nuclear fuel into an annular space defined by a dual tube of inner and outer tubes, the spacer grid taking charge of an important function of inhibiting the vibration of the fuel rods caused by the flow of the coolant has no choice but to support only the outer tube of each fuel rod due to its structure. Due to the limitation of this supporting structure, in the case of the inner tube having the slenderness ratio of about 400 or more, only opposite ends of each fuel rod are supported by the upper and lower end plugs. Of course, in the case of the dual-cooled fuel rod, a uranium dioxide (UO2) pellet exists between the inner and outer tubes. Thus, the vibration of the inner tube is expected to be inhibited to a certain extent. However, in cases of a fuel rod having an slenderness ratio of about 400 or more, it is easily surmised that a vibration amplitude of the inner tube is remarkably great, as compared to the outer tube having numerous support points formed in an axial direction of the fuel rod by the spacer grid. The flow of the coolant in the internal coolant channel defined by the inner tube of the dual-cooled fuel rod can be interpreted as a problem of vibration of the tube in which fluid flows. Thus, if the vibration of this inner tube is not properly controlled, it is difficult to avoid damage to the fuel rod due to such vibration. Further, there are additional considerations to consider in the case of the application of the dual-cooled fuel rod 10. In order to be structurally compatible with the core of the existing pressurized water reactor illustrated in FIG. 2, structural limitations that cannot change positions of core structural components, for instance guide thimbles 140, in the nuclear fuel assembly 100 must be accepted, and the outer diameter of the fuel rod yet must still be increased. As such, an gap between the dual-cooled fuel rod and the spacer grid has no alternative but to be considerably narrowed compared to an existing gap. For example, as in FIG. 3, if the nuclear fuel assembly is configured so that the positions of the guide thimbles 140 are maintained without change according to a design of the nuclear fuel assembly forming a 12×12 array, the gap between the dual-cooled fuel rod 10 and the unit spacer grid is reduced from 1.45 mm, which is the existing gap, to about 0.39 mm. Thus, due to the narrow gap between the dual-cooled fuel rod 10 and the unit spacer grid, technology that forms a fuel rod supporting structure on a surface of the unit spacer grid in order to support the fuel rod has until now been difficult to apply to the dual-cooled fuel rod. Accordingly, the present invention has been made keeping in mind the above problems, namely, that it is difficult for a support structure, such as a conventional spacer grid formed of a thin plate having grid springs and dimples, to support the dual-cooled fuel rod, and thus embodiments of the present invention provide a perforated plate support, in which a thin plate is provided with numerous holes, and each end plug of the dual-cooled fuel rod is coupled to the thin plate around each hole by, for instance, welding, thereby effectively supporting the fuel rod although an gap between the fuel rods is narrow. Further, there is provided a perforated plate support capable of coupling two of at least two dual-cooled fuel rods into which a dual-cooled fuel rod having a great slenderness ratio is segmented. According to an exemplary embodiment of the present invention, there is provided a perforated plate support, which supports dual-cooled fuel rods, each of which has concentric outer and inner tubes and is coupled with upper and lower end plugs at upper and lower ends thereof, and guide thimbles, each of which is used as a passage for a control rod, and which is formed as a support plate having the shape of a flat plate, which includes internal channel holes, each of which has a diameter corresponding to an outer diameter of the inner tube, guide thimble holes, each of which has a diameter corresponding to an outer diameter of the guide thimble, and sub-channel holes around each internal channel hole. The upper or lower end of the dual-cooled fuel rod is coupled to the support plate such that the outer diameter of the inner tube is matched with the diameter of the internal channel hole. The dual-cooled fuel rod may be segmented into at least two dual-cooled fuel rods. Further, the first of the dual-cooled segmented fuel rods may be located on the support plate and be coupled to the internal channel hole at a lower end thereof, whereas the second dual-cooled segmented fuel rod may be located under the support plate and be coupled to the internal channel hole at an upper end thereof. The upper or lower end of the dual-cooled fuel rod may be welded to the support plate. The internal channel holes may be arranged in such a manner that lines extending from centers of the neighboring ones of the internal channel holes form square grids, and the sub-channel holes may be disposed in centers of the square grids, respectively. At this time, each sub-channel hole may have a diameter identical to the outer diameter of the outer tube. The internal channel holes may form a 12×12 array. The support plate may overlap with another support plate so as to form two layers coupled up and down. The support plate may include steps around each internal channel hole. Each step may have arcuate contact faces so as to correspond to the outer diameter of the outer tube. The outer tube may come into close contact with the steps and be welded to the steps. At this time, the inner tube may protrude outwardly from the upper or lower end of the dual-cooled fuel rod, and the protruding inner tube may be inserted into the internal channel hole. The support plate may be coupled with an outer plate along an outer circumference thereof which has guide taps and flow mixture vanes alternately formed. At this time, the outer plate may be coupled to the support plate so as to protrude from one of upper and lower surfaces of the support plate. The support plate may include two support plates, from one surface of which the outer plate does not protrude, and which are coupled to each other. According to embodiments of the present invention, the perforated plate support can properly support the fuel rods even if an gap between the fuel rods is reduced to about 1.0 mm or less because a nuclear fuel assembly is constituted of dual-cooled fuel rods, each of which has a dual tube structure of inner and outer tubes and is subjected to an increase in outer diameter. This perforated plate support has the shape capable of supporting the fuel rods despite the narrow gap between the fuel rods, and makes it possible to support the inner tube, each of which has difficulty in forming a support section except an upper or lower end plug, at regular gaps, unlike the outer tube supported outside at regular gaps, so that the fuel rod is inhibited from excessively being vibrated by a coolant flowing through an internal channel of the fuel rod, and a possibility of fretting wear occurring due to either friction between the fuel rod and the support section or impact load is remarkably reduced. As a result, it is possible to prevent operational efficiency of a nuclear power plant from being reduced by damage to the fuel rod. Further, the fuel rod is supported in a transverse direction rather than in an axial direction as in the prior art, so that, although impact load against the fuel rod occurs in the transverse direction, i.e. on the side, of the fuel rod, excessive plastic deformation is inhibited. Thus, a nuclear fuel assembly, impact strength of which is promoted, can be obtained. This means that, even in the case in which the design peak ground acceleration in the seismic design codes based on earthquakes and accidents involving the loss of coolant is boosted to 0.3 G, more excellent structural soundness can be ensured by securing sufficient impact strength. In detail, in the case of the convention spacer grid, thin plates are arranged in a lengthwise direction of the fuel rod, and then are welded to each other. As such, there is a high possibility of, when the impact load is applied on the side, excessive plastic deformation occurring to limit a path along which a control rod is inserted. However, in the case of the perforated plate support, the plate is disposed in a transverse direction rather than in a lengthwise direction of the fuel rod. Thus, when the impact load is applied on the side, the perforated plate support makes a rigid motion along with the fuel rod. As such, the impact strength is considerably increased compared to the conventional spacer grid. Thus, although the design peak ground acceleration against the seismic load is boosted, more excellent structural soundness can be obtained. Reference will now be made in greater detail to an exemplary embodiment of the invention with reference to the accompanying drawings. According to an embodiment of the present invention, the detailed configuration of a perforated plate support 100 is well shown in FIGS. 4 and 5. The perforated plate support 100 is used to support dual-cooled fuel rods 10, each of which has concentric outer and inner tubes 12 and 14 and is coupled with upper and lower end plugs 16 and 18 at upper and lower ends thereof, and guide thimbles 20, each of which is used as a passage for a control rod (not shown). Particularly, the perforated plate support 100 is formed as a support plate 110 having the shape of a flat plate, which includes a plurality of internal channel holes 112 having a diameter corresponding to an outer diameter of the inner tube 14, and a plurality of guide thimble holes 114 having a diameter corresponding to an outer diameter of the guide thimble 20. The support plate 110 includes sub-channel holes 116 around each internal channel hole 112. The upper or lower end of the dual-cooled fuel rod 10 is coupled to the support plate 110 such that the outer diameter of the inner tube 14 is matched with the diameter of the internal channel hole 112. Here, the internal channel hole 112 is for a coolant flowing in the inner tube 14 of the dual-cooled fuel rod 10, whereas the sub-channel hole 116 is for a coolant flowing out of the outer tube 12 of the dual-cooled fuel rod 10. Thus, the support plate 110 has two channels for dually cooling the inside and outside of the dual-cooled fuel rod 10. The dual-cooled fuel rod 10 may be configured as a single fuel rod. However, the dual-cooled fuel rod 10 is preferably configured as at least two segmented fuel rods, because support soundness of the dual-cooled fuel rod 10, particularly the inner tube 14, is promoted. If the dual-cooled fuel rod 10 is configured of a single fuel rod, the perforated plate support 100 can be coupled only to two places, i.e. the upper and lower ends, of the dual-cooled fuel rod. In contrast if the dual-cooled fuel rod 10 is segmented into two or more dual-cooled fuel rods, the first dual-cooled segmented fuel rod 10′ located on the support plate 110 is coupled to the internal channel hole 112 at a lower end thereof and the second dual-cooled segmented fuel rod 10″ located under the support plate 110 is coupled to the internal channel hole 112 at an upper end thereof As such, the number of points supporting the inner tube 14 is increased. The fact itself that the number of points supporting the inner tube 14 is increased can be sufficiently anticipated improving the support soundness of the dual-cooled fuel rod 10. Furthermore, since the slenderness ratio of the dual-cooled fuel rod 10 is reduced in inverse proportion to the number of segmented fuel rods, rigidity against the flow-induced vibration is greatly increased as well. When the support plate no is welded to the upper or lower end 12″ or 12′, respectively, of this dual-cooled fuel rod 10 (including the dual-cooled segmented fuel rods 10″ and 10′, respectively), this makes it possible to secure more sufficient coupling strength. Meanwhile, the internal channel holes 112 are arranged in such a manner that lines extending from the centers of the neighboring ones of the internal channel holes 112 form square grids. This arrangement can be regarded as the configuration of a typical nuclear fuel assembly. In this embodiment, the internal channel holes 112 are arranged so as to conform with a 12×12 array. Of course, the guide thimble 20, which is used as the passage for the control rod and is supported in the guide thimble hole 114, is coupled with an end fitting at a lower end thereof so that there is no flow of the coolant. At this time, as illustrated in FIG. 4, the sub-channel hole 116 is formed in the center of the square grid. With this configuration, four of the sub-channel holes 116 enclose one of the internal channel holes 112 in a square shape. Here, the diameter of the sub-channel 116 can be identical to the outer diameter of the outer tube 12. Further, the sub-channel holes 116 are provided around the guide thimble hole 114. Since the outer diameter of the guide thimble 20 is typically larger than that of the dual-cooled fuel rod 10, the sub-channel holes 116 provided around the guide thimble hole 114 partially overlap with the guide thimble hole 114 when each sub-channel 116 has the same diameter as the outer tube 12 as illustrated in FIGS. 4 and 5. Further, the support plate 110 may be formed in such a manner that two support plates having the same geometry overlap with each other so as to be coupled up and down. In this case, the rigidity of the support plate 110 is further increased. The support plate 110 is provided with steps 118 around each internal channel hole 112, wherein each step 118 has arcuate contact faces so as to correspond to the outer diameter of the outer tube 12. The arcuate contact faces of each step 118 come into close contact with an outer circumference of the dual-cooled fuel rod 10, particularly an outer circumference of the outer tube 12 of the dual-cooled fuel rod 10, so as to generate sufficient frictional force, thereby improving fixability of the dual-cooled fuel rod 10. Due to the configuration of this step 118, it is possible to eliminate welding when the dual-cooled fuel rod 10 is fixed to the support plate 110. Further, when the steps 118 are formed so as to be exactly symmetrical about the internal channel hole 112, the center of the inner tube 14 of the dual-cooled fuel rod 10 is accurately aligned with that of the internal channel hole 112. Of course, in spite of the configuration of this step 118, the outer tube 12 may come into close contact with the steps 118, and then be welded to the steps 118 for the purpose of firmer coupling. The inner tube 14 may protrude outwardly from the upper or lower end of the dual-cooled fuel rod 10, and then the protruding inner tube 14 may be inserted into the internal channel hole 112. Due to this configuration, the dual-cooled fuel rod 10 can be more accurately aligned with the internal channel hole 112, and coupling strength between the inner tube 14 and the support plate 110 can be improved. The coupled state between the inner tube 14 and the internal channel hole 112 is shown in FIG. 6. If the configuration of the abovementioned step 118 is added to this configuration, the alignment and the coupling strength can be further improved upon. Further, the support plate 110 may be coupled with an outer plate 120 along an outer circumference thereof The outer plate 120 has guide taps 122 for smoothly charging and discharging the nuclear fuel and flow mixture vanes 124 for promoting mixture of the coolant, wherein the guide taps 122 alternate with the flow mixture vanes 124. This outer plate 120 is manufactured by preparing numerous unit outer plates 120′, each of which has the shape of a flat plate in which the guide taps 122 alternate with the flow mixture vanes 124, welding four of the unit outer plates 120′ to the outer circumference of the support plate 110, and welding ends of the unit outer plates 120′ contacted with each other at the corners of the support plate 110. At this time, the outer plate 120 may be coupled to one of the upper and lower surfaces of the support plate 110, and then two support plates, from one surface of which this outer plate 120 does not protrude, may be coupled into one support plate 110. Two surfaces of the two support plates which do not face each other, i.e. which are exposed to the outside, are provided with the abovementioned steps 118. Further, the outer plates 120 contacted when these two support plates are coupled can be welded. Although an exemplary embodiment of the present invention has been described for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims. |
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052689417 | claims | 1. A process for containing radioactive steam emitted from an omega seal between a control rod drive mechanism and an associated adapter tube comprising the steps of; determining an actual diameter of the adapter tube; positioning a canopy seal about the omega seal; and securing said canopy seal to both said control rod drive mechanism and said adapter tube; wherein said canopy seal encloses said omega seal for containing the radioactive steam emitted from the omega seal. 2. The process as defined in claim 1, wherein the step of determining the actual diameter of the adapter tube includes positioning diametrically spaced caliper arms adjacent the adapter tube and sensing the displacement of said caliper arms between a first predetermined position and a second position in contact with the adapter tube. 3. The process as defined in claim 2, wherein said sensor transmits data to a control center which digitally displays said actual diameter of the adapter tube. 4. The process as defined in claim 1, wherein the step of determining the actual diameter of the adapter tube is remotely carried out. 5. The process as defined in claim 1, wherein the step of positioning said canopy seal about the omega seal includes providing a two-piece canopy seal, mounting the two-piece canopy seal in a canopy seal installation fixture, and releasably securing said canopy seal installation fixture to said adapter tube. 6. The process as defined in claim 5, wherein the step of positioning said canopy seal further includes the step of raising said canopy seal into contact with an underside of the control rod drive mechanism after said canopy seal installation fixture is secured to said adapter tube. 7. The process as defined in claim 1, wherein said canopy seal is secured by welding. 8. The process as defined in claim 1, wherein said canopy seal is initially temporarily tack welded in place. 9. The process as defined in claim 1, wherein the area contained by said canopy seal is purged during the step of securing said canopy seal. 10. The process as defined in claim 9, wherein the area is purged with argon gas. |
062529234 | summary | FIELD OF THE INVENTION The present invention relates to monitoring of spent nuclear fuel, and more particularly relates to in-situ, self-powered monitoring of spent nuclear fuel stored in containers. BACKGROUND INFORMATION In the packaging of spent nuclear fuel from nuclear reactors for dry storage or shipment, the primary concern is with the reliable containment of the radioactive nuclear fuel in order to prevent unwanted exposure to radiation. To ensure this protection, the container or canister, and its associated surrounding cask, must meet several criteria. The container must provide containment integrity to prevent the release of radioactive material during normal or hypothetical accident conditions, as well as structural integrity in supporting the weight load of the contained fuel. Thermal protection is required to provide adequate dissipation of the decay heat produced by the spent nuclear fuel, to avoid compromising the integrity of the fuel or the materials of the container or cask. Adequate radiation shielding must also be provided to limit the radiation field external to the package to acceptable values. The container must also protect against the possibility of nuclear criticality, since the contained fuel could represent a critical mass if the system's engineered features were removed or sufficiently degraded, or if the fuel geometry were radically altered through structural degradation. The containers are sealed upon loading, often by double welding, so that re-entry to perform measurements is often not an acceptable option. Instrumentation penetrations are generally regarded as unacceptable because of increased risk of confinement system failure. Because of the uniformity in the manufacture of such systems, monitoring may be performed on a statistically acceptable subset of containers for a given fuel type, to ensure safety of the entire population of containers. Currently, only coarse methods of monitoring are available, for fuel in dry storage. In some metal cask systems, the performance of metallic O-rings is monitored, to ensure seal integrity. In concrete cask systems in which air flow occurs, periodic monitoring of air outlet temperature is conducted. The present invention has been developed in view of the foregoing, and to overcome other deficiencies of the prior art. SUMMARY OF THE INVENTION The present invention provides a method and apparatus which allow remote monitoring of stored spent nuclear fuel, without breeching the containment boundary of the containers in which the spent nuclear fuel is stored. Detectors can be configured as an integral part of a spent fuel container, or as a measurement probe which can be placed adjacent the spent fuel container while it is contained within a storage cask. Monitoring of the neutron and .gamma.-ray fields could be used to indicate the integrity of the fuel or neutron absorber, and the presence of water in the container. Temperature measurements may also be made to determine the temperature profile in the container. Structural degradation or failure of key components within the container may also be detected. An object of the present invention is to provide a system for measuring neutron and .gamma.-ray flux emitted from spent nuclear fuel including a container, spent nuclear fuel in the container, and at least one detector positioned to receive the neutron and .gamma.-ray flux from the spent nuclear fuel. Multiple such "neutron and .gamma.-ray" detectors are preferably positioned in an array either inside or outside the container. The neutron and .gamma.-ray detectors preferably include a semiconductor active region comprising silicon carbide or metal. A cask made of concrete or metal may be placed around the container and detectors to provide shielding, support, and impact protection. Another object of the present invention is to provide a method of measuring neutron and .gamma.-ray flux from spent nuclear fuel, the method comprising the steps of placing spent nuclear fuel in a container, sealing the container, and measuring neutron and .gamma.-ray flux from the spent nuclear fuel with at least one neutron and .gamma.-ray detector. The detector may be sealed inside the container or positioned on the exterior of the container. Another object of the present invention is to provide a method and apparatus for measuring the temperature of spent nuclear fuel in a container. Another object of the present invention is to provide a method and apparatus for sensing the structural integrity of key components of containers for spent nuclear fuel. These and other objects of the present invention will be more apparent from the following description. |
050380478 | abstract | A lead impregnated shielding material hood to protect the head, brain, hypothalamus, and the master glands. A radiation protective shield for preventing direct impingement of radiation rays on the brain, neck, some salivary glands, thyroid and the adjacent body areas of head and eyes and adjacent tissues, of the patient being x-rayed. Having extraoral portion being of shape and size to cover the lead impregnated shielding material, and a substantial area of the extraoral anatomy of the head of the person to thereby intercept rays and prevent direct and/or stray penetration impingement thereon. Having physically protection of the parotid, sublingual and salivary glands. A radiation protective lead impregnated shielding hood shield being to prevent head and eyes from contact with x-rays or secondary x-rays which involve substantially large or intense radiation treatments. |
summary | ||
047711787 | claims | 1. A goniometer stage for an electron microscope, the goniometer stage comprising: a bearing block; a specimen holder rod; a bearing sleeve positioned to accommodate said specimen holder rod therein; a tilting means for tilting said bearing sleeve and said specimen holder rod, said tilting means including: a ball-like inner bearing surface defining a center point and being formed on said bearing block for receiving said bearing sleeve thereon; an outer bearing; a rotatable cylinder rotatably journalled in said outer bearing and defining a tilt axis passing through said center point; adjusting means for adjusting the position of said cylinder and said tilt axis; positioning means for positioning said bearing sleeve and said specimen holder rod about said center point so as to effect a positioning of a section of a specimen; rotation means for imparting rotational movement to said cylinder; and, torsion-resistant means connecting said rotatable cylinder to said bearing sleeve for transmitting said rotational movement to the latter while at the same time permitting said positioning thereof. a housing for accommodating said bearing block; X-direction positioning means for positioning in the X-direction to said tilt axis; a spherical surface formed on said housing and defining a center point coincident with said center point of said inner bearing surface; and, means for adjusting said X-direction positioning means on said spherical surface. 2. A goniometer stage of claim 1, said torsion-resistant means being a torsion-resistant bellows; said positioning means including two mutually separate positioning devices for positioning the tilting about said tilt axis in the Y and Z directions, respectively, independently of each other. 3. The goniometer stage of claim 2, said positioning devices conjointly defining a mechanical-stage guiding arrangement. 4. The goniometer stage of claim 2, said positioning devices being mounted on said rotatable cylinder for engaging said bearing sleeve in the vicinity of said outer bearing. 5. The goniometer stage of claim 1, said bearing sleeve having a ball-like end for engaging said ball-like inner bearing surface, said goniometer stage further comprising: a V-shaped bearing surface formed in said ball-like end for journalling said specimen holder rod; and, spring means for resiliently biasing said rod against said V-shaped bearing surface. 6. The goniometer stage of claim 1 further comprising: 7. The goniometer stage of claim 2, said positioning devices being selected from the group consisting of mechanical drives, piezoelectric drives, pneumatic drives and electric-motor drives. 8. The goniometer stage of claim 2, said tilting means including position transducers mounted on respective ones of said positioning devices for providing an indication of the coordinates of the adjusted specimen section. 9. The goniometer stage of claim 8, said position transducers and said positioning devices being connected to a control arrangement. 10. The goniometer stage of claim 2, said torsion-resistant bellows being made from a metal selected from the group consisting of tombac, bronze and stainless steel. |
description | This invention relates to wireless communications, and more particularly, to wireless communications in networks such as cellular networks. A cellular network is a wireless network that includes multiple cells and base stations. Base stations are towers for mounting antennas, transceivers, and other wireless communications equipment. Each cell is served by one or more respective base stations. Base stations are sometimes referred to as cell sites, cell towers, base transceiver stations (BTS), etc. In a typical cellular network, base stations may be located close to one another (e.g., ¼-½ mile apart) or far away from one another (e.g., 1-2 miles apart). Base station spacing generally depends on the number of active users served by each base station and considerations such as cell site topography. Cellular base stations provide radio coverage for user equipment (UE) such as portable user devices. An example of a portable user device is a mobile phone that is used to make telephone calls. The user device may communicate wirelessly with a neighboring base station that is providing radio coverage. Base stations may collectively provide radio coverage that covers a wide geographic region. The radio coverage of each cell may be approximately circular. The radio coverage of each cell typically overlaps with the radio coverage at least one other cell to provide a continuous region of radio coverage for the cellular network. The amount of overlap between adjoining cells should generally not be too large to minimize waste and to minimize potential interference among the cells. A cellular network implemented using a wireless telephony protocol such as the Global System for Mobile communications (GSM) protocol or a code division multiple access (CDMA) protocol. Regardless of what technology is used, user devices should be able to move from cell to cell (e.g., the wireless technology should support a handover mechanism) without losing wireless connectivity. User devices such as mobile phones are typically designed to operate over a range of possible transmit powers. Wireless communications systems use transmit power control (TPC) commands to control the transmit powers of user devices in the system. Consider a first scenario in which a user device is transmitting at maximum power levels to a base station. If there are other devices that are also communicating with the base station, the signal transmitted by the user device may be too strong and may interfere with the signals transmitted by the other devices. If the full power of the devices is not required for satisfactory communications, the base station may send a TPC command that instructs the user device to transmit at a lower power level. By reducing transmit powers whenever possible, interference can be reduced and the number of simultaneous users can be maximized. Reduction in transmit power may be common in urban regions that have dense populations of active user devices and short distances between cell towers. In contrast, consider a second scenario in which a user device is transmitting at a low power. If the transmit power is too weak, the base station may not be able to properly receive the transmitted signal from the user device. In this scenario, the base station may send a TPC command that tells the user device to transmit at a higher power level. Operation at higher powers may be common in suburban regions that have sparser populations of active user devices relative to urban regions. Because not as many user devices are simultaneously communicating with a given base station and because cell signals must travel greater distances, user devices may transmit at higher powers. Because of the differences in topology and cell usage patterns in different geographic locations, cellular telephones tend to operate at different powers in different environments. For example, cellular telephones in urban environments may tend to operate at lower transmit powers than cellular telephones in suburban areas. Each different type of operating environment may, in general, be characterized by its own unique set of cellular telephone transmit power statistics. In conventional networks, cellular telephones are provided with a single set of transmitter settings. These settings dictate how the cellular telephone transmits cellular signals under different conditions. For example, the transmitter settings may dictate that low transmit powers should be produced by activating a single radio-frequency power amplifier gain stage in the cellular telephone, whereas high transmit powers should be produced by activating two radio-frequency power amplifier gain stages. Although use of a generic set of transmitter settings for all cellular telephones in a cellular telephone network may be straightforward to administer, inefficiencies and performance issues may arise during operation. In particular, the generic transmitter settings might result in high call quality in suburban environments or other environments where transmit powers tend to be high, but may result in suboptimal call quality in urban environments. As another example, the generic transmitter settings might result in excellent transmitter power efficiency performance in urban environments, but might exhibit lower than desired transmitter power efficiency performance in suburban environments. Because call quality and power consumption must be acceptable for all users, the generic transmitter settings that networks use represent a compromise. Settings are not optimized for either urban or suburban environment, but rather are selected so that adequate performance is obtained on average. Although use of generic transmitter settings is generally satisfactory, recent advances in cellular telephones are placing increasing demands on cellular networks and cellular telephone manufacturers. Increased cellular telephone usage is giving rise to a greater potential for interference between devices. Battery life is also of growing importance, because users are using their cellular telephones to browse the Internet and handle other power-intensive tasks. It would therefore be desirable to be able to provide ways in which to optimize the performance of wireless networks and wireless electronic devices such as cellular telephones. Electronic devices such as portable user devices may communicate wirelessly with a nearby base station. A user device may include storage and processing circuitry, a transceiver, a power amplifier, a voltage supply, Global Positioning System (GPS) unit, etc. The user device may be used to transmit wireless signals with a certain transmit power. The transceiver may be adjusted to change the transmit power. The power amplifier may have a varying number of amplifying stages and may be biased at a tunable bias voltage. The power amplifier may be tuned to change the transmit power as well. Each user device may keep a log that records the transmit power at different points in time. Each data point may also be tagged with the current corresponding location of the user device. The current location of the user device may be determined by the GPS unit, a base station identifier, a wireless hotspot identifier, etc. A user may dock his user device and may connect the user device to a service account. The user device may upload its log to the service account. The uploaded logs of each user device may be aggregated to a central power optimization server. The power optimization server may continuously pull and store the logs of user devices from different geographical regions. The power optimization server may organize the data into separate groups corresponding to the different regions (e.g., sorting the data according to the location tags). A region may be any desired geographical area (e.g., a city, a state, a country, etc.). Furthermore, the data may be organized according to the type (model) of user devices. The power optimization server may calculate a cumulative distribution function (CDF) of transmit power based on the logs retrieved for each region and user device model. The CDF may be used to show the percentage of time that a typical user device spends transmitting wireless signals at various power levels. An optimum transmit power setting may be determined for each region and model type based on the CDF. The optimum transmit power setting may be a setting that is optimized to transmit power at a power level that corresponds to the peak of the CDF. A set of optimum transmit power settings may be calculated in this way for any number of regions. At regular time intervals (e.g., once every few months), the power optimization server may update the optimum power settings based on the old optimum power settings and data gather from the newly retrieved logs since the previous update. A user may download the updated power settings by docking his device or by retrieving the optimum settings through a wireless network. A user may download the optimum settings for the entire world or may download a subset of the optimum settings that corresponds to the region in which the user resides. When the user device is in use, the user device may calculate its current location using the base station identifier, for example. Based on the current location, the user device may operate at the corresponding optimum transmit power settings. A user device configured in this may be more power efficient. It is important to note that once the user moves to another region, the optimum transmit power settings may be changed to reflect the settings that are suitable for the new location. In a scenario in which a user device has downloaded a subset of optimum settings that does not include the optimum settings for a given location, the user device may still be able to retrieve the appropriate transmit power settings wirelessly. Further features of the user device, its nature and various advantages will be more apparent from the accompanying drawings and the following detailed description. This relates to cellular networks and wireless electronic devices such as cellular telephones that communicate within cellular networks. A cellular network is a radio network that is made up of a group of cells. Each cell includes a base station that provides wireless coverage for that cell. A base station may be a tower. Antennas, transceivers, and other wireless communications equipment may be mounted on the tower to provide wireless communications capabilities. A base station may therefore be referred to as a cell tower, a cell site, a base transceiver station (BTS), etc. A base station may be located at the center of each cell. Using a network of base stations, a cellular telephone operator may provide wireless coverage for cellular telephones over a wide area. Multiple cells may be located throughout a geographic region to provide wireless coverage for that geographic region. The wireless coverage of each cell may partly overlap with the wireless coverage of at least one other cell to provide continuous or nearly continuous wireless coverage. Each base station can support only a finite number of simultaneous users. As a result, different geographic regions are typically provided with different cell densities. Regions that have more dense concentrations of active users (e.g., urban regions) may require base stations to be placed closer together (e.g., with a smaller radius of coverage for each cell). Regions that have less dense concentrations of active users (e.g., suburban regions) may be provided with base stations that are located farther apart from each other. Cellular carriers typically offer cellular coverage that spans a variety of geographic areas. For example, national cellular carriers offer coverage in urban areas, suburban areas, and rural areas. The characteristics of a cellular network can also vary within areas of these types. For example, some suburban areas may be relatively flat and spread out whereas other suburban areas may be hilly and compact. To enhance operation of a cellular telephone or other portable user device that is wirelessly communicating with a base station in a cellular network, the way in which the device operates can be adjusted as a function of its environment. In particular, the settings of the user device may be optimized so that the device operates differently depending on the characteristics of the cellular network in which the user device is currently located. When, for example, a device is located in a wireless network of a first type (e.g., an urban network), the device may use a first set of settings (e.g., a first set of radio-frequency power amplifier settings). When the device is located in a wireless network of a second type (e.g., a suburban network), the device may use a second set of settings (e.g., a second set of radio-frequency power amplifier settings). Because operation of the device is tailored to the location of the device, performance can be enhanced relative to devices that use the same set of settings in all portions of a network. A cellular telephone or other electronic device such as user device 10 of FIG. 1 may be provided with location-specific settings. User device 10 may be a mobile telephone, computing equipment such as a handheld electronic device, tablet computer, or laptop computer, or other wireless device that communicates with a cellular telephone network. Device 10 may use wireless communications to handle voice traffic and/or data traffic. Device 10 may include storage and processing circuitry such as storage and processing circuitry 12. Storage and processing circuitry 12 may include an applications processor such as applications processor 14, a baseband processor such as baseband processor 16, and storage circuitry such as storage circuitry 18. Additional processors such as digital signal processing chips, application-specific integrated circuits, and other processing components may be included in circuitry 12. Applications processor 14 and baseband processor 16 may be used to perform digital signal processing operations for device 10. For example, applications processor 14 may be used to execute software for implementing a web browser, email application, or other applications. Baseband processor 16 may be coupled to transceiver circuitry such as transceiver 20 and may be used in transmitting and receiving signals over antenna 28. Storage circuitry 18 may include volatile and nonvolatile memory, hard drives, and other storage media and may be used in storing data for the processing components in circuitry 12. Storage and processing circuitry 12 may be connected to a transceiver such as transceiver 20 through data path 19. Storage and processing circuitry 12 may send and receive digital signals to and from transceiver 20 through data path 19. Transceiver 20 may also be controlled by storage and processing circuitry 12 through control line 21. Control line 21 may be used to control the rate at which transceiver 20 is processing data and other transceiver settings. Transceiver 20 may have an input port and an output port. The output port of transceiver 20 may be connected to an input terminal of a radio-frequency power amplifier (PA) such as power amplifier 22. Power amplifier 22 may be used to amplify the radio-frequency signals provided by transceiver 20 prior to transmitting the signals through antenna 28. Power amplifier 22 may have multiple amplifying stages. The number of amplifying stages that are turned on and active may be controlled in real time using control signals applied to power amplifying control line 31. Control line 31 may be connected to an output of baseband processor 16 (or, if desired, the output of other processing circuitry such as applications processor 14). Power amplifier 22 adjusted by control line 31 so that only a single amplifying stage is active or may be adjusted so that two stages are active or so that three amplifier stages are turned on (as examples). It may be desirable to activate most or all of the stages in amplifier 22 in situations in which a large output power is desired. In situations in which less output power is needed, it may be desirable to turn off one or more of the stages of amplifier 22 to conserve power. If the output power from transceiver 20 is sufficient without further amplification, power amplifier 22 may be bypassed (e.g., using switch circuitry). Power amplifier 22 may have an output terminal that is connected to input-output (I/O) circuitry such as I/O circuitry 26. Circuitry 26 may include switching circuitry (e.g., radio-frequency switches), radio-frequency filters (e.g., duplexers), tuning circuits (e.g., impedance matching circuits), etc. For example, circuitry 26 may contain a duplexer that routes signals by frequency. This allows antenna 28 to be shared between transmitter and receiver circuits. The input port of transceiver 20 may be connected to an output terminal of a low noise amplifier (LNA) such as low noise amplifier 24. Low noise amplifier 24 may have an input terminal that is connected to I/O circuitry 26. Low noise amplifier 24 may be used to amplify received signals while generating low noise content. I/O circuitry 26 may include a duplexer. A duplexer is a device that allows for bidirectional communication (e.g., transmitting and receiving wireless signals). I/O circuitry 26 may be connected to an antenna such as antenna 28. Antenna 28 may be used to broadcast and receive analog wireless signals. The duplexer isolates the transmit operation from the receive operation to allow the system to share a common antenna 28. I/O circuitry 26 may include other circuitry such as matching circuits, mixers, etc. Storage and processing circuitry 12 may be connected to a power supply unit such as voltage supply 30. Voltage supply 30 may include biasing circuitry that can be used to provide a controlled (adjustable) output voltage. For example, voltage supply 30 may provide a control voltage VBIAS on bias line 32. Bias line 32 may be connected to a supply terminal of power amplifier 22. The level of VBIAS and the number of active gain stages in amplifier 22 represent adjustable settings that affect the way in which the wireless circuitry (and, more particularly, power amplifier 22 operates). Other wireless settings may also be adjusted (e.g., transceiver settings such as the output power produced by transceiver 20). For clarity, illustrative configurations in which storage and processing circuitry adjusts settings in power amplifier 22 such as the value of VBIAS and the number of active amplifier gain stages are sometimes described herein as an example. This is, however, merely illustrative. Any suitable wireless settings may be adjusted by storage and processing circuitry 12 if desired. During operation of device 10, wireless settings may be adjusted in real time. For example, storage and processing circuitry 12 can turn on and off gain stages in response to required output power levels. VBIAS may also be adjusted depending on current operating conditions. Further optimization may be obtained by adjusting at least some wireless circuitry settings based on information about the position of device 10. For example, storage and processing circuitry 12 may maintain information on the threshold powers that trigger the turning off and on of gain stages. Optimum values for the threshold power settings may be different when device 10 is to be used in an urban environment or a suburban environment. There may be one or more settings of this type in device 10. For example, there may be two, three, four, or more than four settings for which optimum location-dependent values may be ascertained. Although sometimes referred to as location-dependent values, optimum settings may be switched into use based on any suitable variable. As an example, device 10 may select optimum settings to use based on device location, the type of geographic region in which the device is operating (e.g., urban, suburban, rural), the type of topology of the current device location (flat, hilly, mountainous, urban canyons, etc.), other attributes of the current operating zone (i.e., whether or not an area is heavily or lightly populated, whether or not electrical interference is commonly present, etc.). Particularly when the operation of device 10 can be optimized by selecting among sets of available settings such as these, it may be desirable to allow device 10 to determine its current location. Location information can also be inferred (e.g., from the location at which the user of device 10 resides, from information gathered through the cellular telephone network, from postal code information or city name information entered manually by the user, from information that the user sets up in an online database and that is subsequently downloaded into device 10, or from other suitable sources). With one suitable arrangement, device 10 may include a Global Positioning System (GPS) unit such as GPS unit 34 that is used to calculate the current location of device 10. GPS unit 34 may be connected to storage and processing circuitry 12, so that storage and processing circuitry 12 can look up appropriate wireless settings to use in real time, based on the current location data supplied by GPS unit 34. During operation, device 10 transmits wireless signals at a transmit power that is influenced by a number of factors. These factors include the output power of the radio-frequency signals produced by transceiver 20, the number of gain stages that are active in amplifier 22, and the voltage bias VBIAS for amplifier 22 (as examples). As device 10 is used, the link quality with its nearest base station may vary. For example, device 10 may move closer or farther from the base station or the wireless link between device 10 and the base station may become clearer or weaker due to interference from objects or electromagnetic interference sources. The cellular network can issue transmit power control (TPC) commands to regulate the transmit power of device 10 as device 10 communicates with a base station. If the base station senses that device 10 is transmitting with an unnecessarily high transmit power, the base station may send a TPC command to device 10 that directs device 10 to decrement its current transmit power level (e.g., by 1 decibel referenced to one milli-watt (dBm)). On the other hand, if the base station senses that device 10 is transmitting with undesirably weak transmit power, the base station may send a TPC command to device 10 that tells device 10 to increment its current transmit power (e.g., by 1 dBm). TCP commands may increment or decrement the transmit power by any desired incremental amount. An urban region may have a dense population of user devices. The active user devices may all be constantly sending and receiving wireless signals from a corresponding base station. A base station in a dense urban region may not be able to handle a situation in which all mobile phones are transmitting at maximum power. The wireless signals transmitted at such high power may undesirably interfere with one another and may decrease the quality of the signals. The base station may therefore send TPC commands that decrease the transmit power of devices to minimum acceptable levels. These TPC commands can be sent during times when a particularly large number of users are making phone calls or may be sent without regard to traffic levels. FIG. 2 shows a cumulative distribution function (CDF) of transmit power (P_OUT) for an urban region. The CDF illustrates the amount of time a user device spends transmitting at various power levels. From curve 36, one can see that a user device in the urban region may spend a high percentage of the time transmitting at −5 dBm (in this example). It is possible for device 10 to transmit at other power levels even in the urban region, as indicated by the other points on curve 36. For example, during busy hours, the base station may force each user device to operate at a low power level (e.g., less than −5 dBm). Alternatively, during times when users are not making as many phone calls, the base station may allow each mobile phone to transmit at a higher power level (e.g., greater than −5 dBm). Curve 36 can be obtained by logging P_OUT of each user device in a group of user devices operating in the urban area with respect to time. P_OUT with most data points will correspond to the peak of the curve while transmit levels with fewer data points will correspond to the other points on curve 36. The CDF for a group of devices may be fairly different if the devices are used in a suburban region, as shown in FIG. 3. According to curve 38 of FIG. 3, a typical user device in a suburban region may spend more time transmitting at 2 dBm than at any other transmit power levels (in this example). Notice that the peak of curve 38 corresponds to a higher transmit power than the peak of curve 36. As described previously, suburban regions tend to involve relatively more operation at higher power levels, because the density of users is relatively lower in comparison to urban regions and because users are farther from base stations. Curve 38 may also have a local peak at 20 dBm that corresponds to devices 10 operating at the outskirts of the suburban area, as an example. Once the characteristics of operating in a particular region are known (e.g., by collecting data for CDF curves such as curves 36 and 38 of FIGS. 2 and 3), it is possible to optimize wireless settings in a device accordingly. For example, empirical studies and modeling studies may be conducted to determine optimum threshold powers at which gain stages should be turned on and off in power amplifier 22 and optimum voltage bias values for VBIAS. In an urban region (or other type of operating environment), a first set of settings may produce best results for call quality and power consumption efficiency. In a suburban region (or other type of operating environment), a second set of settings may be preferable. Any suitable criteria may be used to determine which settings are optimal. For example, emphasis may be placed on call quality or emphasis may be placed on power conservation. If desired, performance in urban areas may put emphasis on call quality (as an example) whereas performance in suburban areas may put emphasis on battery life (as an example). With one suitable arrangement, optimum performance may be obtained for each operating environment by choosing settings that are optimal when devices are operated at transmit power levels are at the peak of the appropriate CDF graph for that type of environment. For example, an optimal set of urban settings can be identified based on the assumption that urban devices will tend to operate at output power levels of −5 dBm, whereas an optimal set of suburban settings may be identified based on the assumption that suburban devices will tend to operate at higher power levels (see, FIG. 3). User device transmit power settings may be adjusted. Each unique transmit power setting may correspond to a different power transfer function (e.g., transmit output power P_OUT versus input power P_IN), as shown in FIG. 4. Consider a first transfer function that includes curves 40-1 and 40-2. The first transfer function may have a first amplifier gain switch point at 1 dBm (e.g., the point at which the first transfer function is discontinuous). Point A on the first transfer function may correspond to a desired operating point optimized for operation in an urban region. Point A may correspond to a P_OUT of −5 dBm, for example. Point A may also be located adequately away from the first amplifier gain switch point. A VBIAS of 2 volts (V) may be used to bias power amplifier 22 to obtain the first transfer function. Control path 31 may be used adjust the number of amplifying stages in power amplifier 22 and to tune P_IN (e.g., to adjust the operating point along curve 40). For example, curves 40-1 and 40-2 may correspond to operating points generated by one and two amplifying stages in power amplifier 22 respectively. Thus, point A only requires power amplifier 22 to use a single amplifying stage, because point A sits on curve 40-1. The first transfer function in this example is not necessarily optimized for use in a suburban region. For operation in suburban environments, a second transfer function may be used. The second transfer function may include curves 42-1 and 42-2 of FIG. 4. The second transfer function may have a second amplifier gain switch point at −3 dBm (as an example). The second transfer curve may be optimized to transmit power at a higher level that is suitable for suburban regions. For example, point B on the second transfer curve may correspond to a desired operating point suitable for operation in a given suburban region. Point B may correspond to a P_OUT of 2 dBm. Curves 42-1 and 42-2 may correspond to operating points generated using either one or two amplifying stages in power amplifier 22, respectively. Because point B is located to the right of the second amplifier gain switch point, point B corresponds to a situation in which two amplifying stages are being used. The use of two amplifying stages instead of a single stage may allow a lower VBIAS of 1.5 V to be used (a lower VBIAS shifts a transfer function vertically downwards). The first and second transfer functions may correspond to optimum power settings for typical urban and suburban users, respectively (as an example). The appropriate power settings may be obtained by adjusting the number of amplifying stages in power amplifier 22, adjusting the amplifier gain switch point, tuning VBIAS, adjusting the settings of transceiver 20, etc. Once a device is configured to operate with a given set of desired optimum power settings, the transfer function of the device may remain fixed until a new set of power settings is configured. The actual transmit power of the device, however, may be adjusted by TPC commands that move the operating point of the device along the fixed transfer function. For example, once device 10 is configured with an urban power setting described in connection with FIG. 4, the amplifier gain switch point may be fixed at 1 dBm. The operating point of device 10 may, however, move along curves 40-1 and 40-2. TPC commands may only require device 10 to transmit power along curve 40-1 (as an example). Optimum transmit power settings may be obtained by aggregating logs of operating data from user devices, as shown in FIG. 5. FIG. 5 illustrates a simplified example in which two types of users are located in two separate regions. Urban users are located in an urban region and suburban users are located in a suburban region. Each user device may store a log. The log may be stored using storage 18 in each device 10. The log may record the transmit power level P_OUT at different points in time (e.g., t1, t2, t3, etc.). The transmit power of the urban user devices may vary about an average power of −5 dBm while the transmit power of the suburban user devices may vary about an average of 2 dBm, as an example. Each log may also keep track of the current location of the user device by analyzing the cell identifier broadcasted by the closest base station, by using the built-in GPS unit of device 10, or by using other location or region-based information. If desired, the quality (e.g., signal integrity) of a call may also be logged to help provide additional information. Signal quality may be measured using any suitable metric such as bit error rate, the frequency of dropped calls, signal strength (e.g., “bars” of strength), etc. The logs of each user device may be aggregated and analyzed. Data analysis operations of this type may be performed using computing equipment. For example, these aggregation and analysis operation may be performed by a power optimization server during the operations of step 44 of the flow chart in FIG. 5. The power optimization server may compile a cumulative distribution function (CDF) for each separate region (see, e.g., FIGS. 2 and 3). Regions types may be identified in advance (e.g., urban and suburban) or the analysis operations of step 44 may be used in identifying suitable region types (e.g., hilly rural-suburban, sparse urban, etc.). Different transmit power settings may be identified that are optimized for each different region. It may be desirable to have transmit power settings that ensure optimum performance when a device is operating at the peak of the CDF for each region or performance can be operated over the entire range of possible transmit powers using the CDF as a weighting function (as examples). Power settings 46 and 48 may be determined by analyzing the aggregated information during step 44. Power settings 46 that are supplied by the operations of step 44 may be the optimal transmit power settings for typical urban users. Power settings 48 that are supplied by the operations of step 44 may be the optimal power settings for typical suburban users. Power settings 46 and 48 may correspond to power transfer functions of the types described in connection with FIG. 4, as an example. This is merely an illustrative example. The aggregation of wireless usage data from logs of different users in this way may be used to determine optimal power settings for users that are located in more than two different regions. The regions can be any desired size and may be selected using any suitable classification criteria. For example, optimal settings may be provided on a cell to cell basis, a city to city basis, a country to country basis, regions chosen based on general geography, etc. Illustrative steps involved in performing the operations of step 44 of FIG. 5 are shown in the flow chart of FIG. 6. At step 50, each user device logs the time spent at each transmit power level. Power level and time data may be maintained in storage 18 by storage and processing circuitry 12. Each data point may be tagged with the current region of the device. The current region in which the device is located may be determined from the base station identifier that is broadcasted by a neighboring base station or may be determined using GPS unit 34 (as examples). If desired, each user device may compile a CDF internally as the log is being update. At step 52, a user may dock device 10 to connect the device to an online service such as a user device service account. The service may be implemented using a local application (e.g., an application running on a user's personal computer) and a remote application (e.g., a server application with which each local application communicates over the Internet. Once logged data from numerous users and geographic locations has been obtained and analyzed, the service may generate recommended settings to use in each region. The service may then be used to provide appropriate settings to device 10. During the process of uploading log data from device 10 to the service, the service may pull the log from the phone and forward the log to a power optimization server (step 54). The power optimization server may continuously receive and store incoming logs as each user docks device 10. The power optimization server may be implemented using one or more computers that are located at one or more geographic locations. The power optimization server and the service that is implemented on the power optimization server may be part of a larger service (e.g., a media download service) or may be associated with other services to which the user can connect device 10. In addition to retrieving logs and distributing optimum transmit power settings, the service may provide a medium through which music, games, applications and other items may be purchased or downloaded. The service may, for example, include the functions of the iTunes® digital media player application of Apple Inc. of Cupertino, Calif. The power optimization server may aggregate and organize the data (i.e., the information in each log). Log data may be organized using predetermined regions or regions can be identified during the organization process. Data may be divided into cities, cell sites, countries, or any other suitable region size or type (step 56). At step 58, the data may be further segmented according to the type (model) of device 10. The power optimization server (or other computer-implemented power optimization service) may determine an optimized transmit power setting for each type of device in each region. The power optimization server may calculate the optimized power settings based on a combined CDF of all the aggregated logs for each type of user device in each region. The optimum transmit power settings may correspond to the peaks of the combined CDFs (as an example). In operation, the transmit power of device 10 will vary depending on the TPC commands, but while varying, device 10 will generally be operating efficiently about the optimum transmit power setting for a certain region according to a transfer function that is specified by the optimum transmit power settings. Optimum settings may be stored on an online server or other computing equipment for subsequent downloading by users. Optimum settings may be downloaded over a wireless network (e.g., periodically, in real time, or whenever a user enters a new region), may be built into devices, and/or may be downloaded into devices when a user connects a device to a service (e.g., when syncing a device to a personal computer on which a client portion of a service is running). FIG. 7 is a flow chart of illustrative steps involved in downloading the optimum transmit power settings onto user device 10. At step 60, device 10 may download in advance the optimum power settings while it is docked and is connected to the online service. Alternatively, device 10 may retrieve the optimum transmit power settings over a wireless network (e.g., when no connector port is available or when it is desired to receive real-time updates to the settings as the user travels between different regions of a network). Device 10 may download the optimum power settings for all the regions in the world or for all regions in a national network. If desired, device 10 may only download a subset of the optimum power settings to save memory. For example, a user who is a resident of California may only need to download a subset of the optimum transmit power settings optimized for a default region such as California, a metropolitan region in California, the North American continent, etc. While device 10 is in use, device 10 may determine the region in which it is currently located (step 62). Device 10 may determine its current location by using built-in GPS unit 34, by using the base station identifier that is broadcasted by a neighboring base station, by using a local wireless hotspot identifier, etc. After determining its current location, the user device may operate using the optimum power settings that correspond to its current location (step 64). In a scenario in which the phone only downloaded a subset of optimum power settings and the subset of optimum power settings does not contain the desired setting for the current location, the user may retrieve the desired power settings via a wireless network. For example, the user who is a resident of California may only download the optimum power settings for California or for the North American continent. The user may travel to China. The subset of optimum power settings that is initially resident in the user's device 10 does not include China (in this example). As a result, the user may have to download the optimum power settings corresponding to the Asia region via a wireless network. If desired, the user may dock his phone and select to download the desired optimum settings corresponding to regions other than the default region. This process may loop back to step 62 as indicated by path 66. The power optimization server may constantly be collecting and aggregating data. The power optimization server, however, may only analyze and update the optimum power settings once every few months based on the current optimum transmit power settings and the newly collected data (i.e., user logs). The user may be notified when he docks the phone that new optimum settings are available for download. If desired, the optimum settings may be selected based on information associated with the user (e.g., an associated region). For example, a user may provide the service with his current residence address or postal code to denote the associated region. Alternatively, the service may have a record indicating the associated region in which the phone was purchased or registered. The service may simply provide the device with an optimum transmit power setting that corresponds to the user's address, the region of registration, etc. If desired, the user device may download a complete lookup table (LUT) that has the optimum power settings for each corresponding postal code or region. The LUT may be stored in storage circuitry 18 (as an example). FIG. 8 shows how a user device such as a cell phone may communicate with the online service. Device 10 may be docked to connect to a personal computer such as personal computer 68. Personal computer 68 may be running a client portion of the service. Device 10 may be connected to personal computer 68 through a connector such as connector 70. Personal computer 68 may be part of a computer network that is connected to the Internet. Personal computer 68 may be connected to the Internet using physical wiring (e.g., Ethernet technology) or using a suitable wireless communications protocol (e.g., Wi-Fi®). Computing equipment such as computing equipment 72 may also be connected to the Internet. Computing equipment 72 may be running the service that can be used to update (sync) various user device settings. Device 10 may be docked in this way to send data (e.g., transmit power logs) to computing equipment 72 through the Internet. Computing equipment 72 may also send optimum transmit power settings, the latest carrier settings, and other information to the docked device through the Internet. If desired, device 10 need not be docked and may communicate directly with computing equipment 72 (as indicated by wireless propagation path 74 in FIG. 8). The foregoing is merely illustrative of the principles of this invention and various modifications can be made by those skilled in the art without departing from the scope and spirit of the invention. |
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054467730 | claims | 1. A heterogeneously loaded fast reactor core comprising a plurality of core fuel assemblies and a plurality of target fuel assemblies, each of said core fuel assemblies comprising a hexagonal wrapper tube and a plurality of core fuel rods positioned within said wrapper tube, each of said core fuel rods having a first predetermined diameter, each of said target fuel assemblies comprising a hexagonal wrapper tube and a plurality of target fuel rods positioned within said wrapper tube, each of said target fuel rods having a second predetermined diameter, each of said target fuel rods containing 20 to 50% of minor actinide nuclides, said second predetermined diameter being smaller than said first predetermined diameter, each of said wrapper tubes of said core fuel assemblies and of said target fuel assemblies, being of the same shape and size, each of said core fuel assemblies having 271 core fuel rods, each of said target fuel assemblies having either 331 or 397 target fuel rods, said fast reactor core having 30 to 50 target fuel assemblies heterogeneously dispersed and loaded therein. each of said core fuel assemblies comprising a hexagonal wrapper tube and a plurality of core fuel rods positioned within said wrapper tube, each of said core fuel rods having a first predetermined diameter, each of said target fuel assemblies comprising a hexagonal wrapper tube and a plurality of target fuel rods positioned within said wrapper tube, each of said target fuel rods having a second predetermined diameter, each of said target rods containing aluminum oxide, magnesium oxide, aluminum magnesium oxide, cerium oxide, gadolinium oxide, aluminum nitride or zirconium nitride, to which 20 to 50% of minor actinide nuclides are added. said second predetermined diameter being smaller than said first predetermined diameter, each of said wrapper tubes of said core fuel assemblies and of said target fuel assemblies, being of the same shape and size, each of said core fuel assemblies having 271 core fuel rods, each of said target fuel assemblies having either 331 or 397 target fuel rods, said fast reactor core having 30 to 50 target fuel assemblies heterogeneously dispersed and loaded therein. 2. The fast reactor core as claimed in claim 1, wherein the basic material of the fuel containing minor actinide nuclides is a mixed oxide or nitride of uranium and plutonium. 3. A heterogeneously loaded fast reactor core comprising a plurality of core fuel assemblies and a plurality of target fuel assemblies, |
description | This application is the US national phase of international application PCT/IT03/00049 filed 03 Feb. 2003, which designated the U.S. and claims priority to IT Application No. RM02A000071 filed 11 Feb. 2002. The entire contents of these applications are incorporated herein by reference. The invention described herein relates to a container for a vial of radiopharmaceutical as well as a set for the infusion of the radiopharmaceutical from the vial housed in the container into a patient or for the transfer of the radiopharmaceutical elsewhere. Currently, radiopharmaceuticals, and particularly but not exclusively, those containing beta-emitting radioisotopes generally destined for infusion into patients, are contained in vials for intravenous injection, equipped with a hermetically sealed rubber cap through which the needle of a syringe is inserted for the extraction of the radiopharmaceutical to be injected or for its transfer elsewhere to a different receptacle, Traditionally, the radiopharmaceutical vials are in turn housed in a lead container. This type of radioprotection using lead containers presents many drawbacks both from the point of view of storage and transportation of the radiopharmaceutical and from that of its subsequent handling for use. Lead containers are heavy, a factor which has a substantial adverse effect on the transportation and storage of the radiopharmaceutical. What is more, owing to their opacity, lead containers prevent visualisation of the contents of the radiopharmaceutical vial. The operator, in fact, has to open them to check their contents and state of conservation, check for any breakage of the vial with a major risk of contamination, and, if required, check the dose of radioactivity. Moreover, in the administration of a radiopharmaceutical to a patient or when transferring it to another receptacle. The operator handling it or aspirating it with a syringe or some other device risks receiving a dose of radiation even as a result of contact with the radiopharmaceutical itself. Another by no means negligible problem in intravenous infusion is that of accurately measuring the amount of radioactive substance infused. This problem was addressed, for example, in U.S. Pat. No. 5,529,189 granted to Feldschuh on Jan. 25, 1996. The aim of that patent was to provide a disposable set for administering a precise dose of radioactive substance to a subject with an accuracy of at least 99.9% by weight. Nevertheless, even if this objective is effectively achieved, the fact remains that according to the above-cited patent the vial of radioactive substance has to be handled with great care owing to the substantial risk to the operator. One of the objectives of the invention described herein is therefore to provide a container for vials of radiopharmaceutical made of a material capable of shielding the operator from radioactive emissions, and particularly beta-emitting isotopes. Another objective of the present invention is to provide an easily manageable, light-weight container. Yet another objective of the present invention is to provide a container for vial of radiopharmaceutical that enables the contents to be identified without needing to open it. Another objective of the present invention is to allow the shipment and transportation of precalibrated, customised radiopharmaceuticals for individual patients in containers in which the radiopharmaceutical can be checked by the operator as corresponding to the dosage amount desired. Yet another objective of the present invention is to allow the infusion of the radiopharmaceutical in a patient or its transfer elsewhere without any need for handling the vial of radiopharmaceutical. One initial aspect of the present invention aims at achieving the above-mentioned objectives by providing a container for vial of radiopharmaceutical made from a material suitable for shielding the operator from the radiation emitted by the radiopharmaceutical through the vial and consisting of a receptacle with a cavity capable of containing the vial of radiopharmaceutical and of a lid coupled to the receptacle for closing the container, said lid being equipped with a central through-hole. One initial additional objective of the present invention is to allow infusion of the radiopharmaceutical in a patient or its transfer elsewhere without any need to aspirate the radiopharmaceutical with syringes in order to extract it from the vial. A second additional objective of the present invention is to allow accurate measurement of the amount of radiopharmaceutical infused in a patient or transferred elsewhere to a different receptacle by reading its volume. A second aspect of the present invention aims at achieving the above-mentioned additional objectives by providing a set in combination with the above-mentioned container housing the radiopharmaceutical vial and consisting of: a saline solution bottle containing saline solution; an infusion catheter equipped with twin connectors, one for inserting a needle into the bottle of saline solution and a second connector for a second needle, inserted, via the central through-hole in the lid, into the cap of the vial of radiopharmaceutical in such a way as not to be immersed in the radiopharmaceutical; a second infusion catheter equipped with twin connectors, one for the insertion of one needle, via the through-hole in the lid, into the cap of the vial of radiopharmaceutical, and the other for a second needle inserted in the patient's vein or elsewhere, the first needle of this second catheter being long enough to touch the bottom of the vial of radiopharmaceutical. With reference to the drawings, FIGS. 1 and 2 show the radiopharmaceutical vial container according to the invention, partly in section, partly in side view, and from above, respectively. It consists of receptacle 1 and lid 2. A radiopharmaceutical vial for intravenous infusion is represented in FIG. 1 with dashed lines and is marked 3. The radiopharmaceutical vial 3 is traditionally a cylindrical UNI 6255 pressed glass vial, or other similar receptacle conventionally used for the same purpose, with an externally enlarged wide mouth 30 on which a rubber cap (not shown) is hermetically sealed with an aluminium crimp-cap seal. Vial 3, e.g. a 20 ml vial, has a cylindrical wall 31, a bottom 32 and a portion 33 widening downwards from mouth 30 to cylindrical wall 31. The radiopharmaceutical to be contained in the vial is a beta-emitting isotope, such as 90Y-biotin, 90Y-DOTATOC, 90Y-MoAbs amongst others. Receptacle 1 is preferably cylindrical and has a cavity 10, which is also cylindrical, capable of containing radiopharmaceutical vial 3 with a mobile coupling. That is to say, it is preferable that the diameter of cavity 10 should be slightly larger than the outside diameter of wall 31 of cylindrical vial 3 so that the latter, which rests on the bottom 11, is prevented from making excessive radial movements and consequently knocking against vertical wall 12 of receptacle 1. In its upper part cavity 10 widens into compartment 13 of greater diameter whose inner wall presents a threaded portion 14. As can be seen in FIG. 1, the height of cavity 10 is such that the vial projects with its mouth 30 beyond the upper rim of vertical wall 12 of receptacle 1. Lid 2 is screwed onto receptacle 1 to close the container. Lid 2 is likewise cylindrical and advantageously is formed in one piece from an upper disk 20 of the same diameter as receptacle 1. The upper disk 20, the rim of which presents a milled or knurled edge 21, to enhance the tightness of fit of lid 2, extends downwards in a similar cylindrical portion 22, with a diameter measuring less than that of the upper disk. The size of cylindrical portion 22 is such that it fits into compartment 13 of receptacle 1 of smaller diameter. Cylindrical portion 22 presents an outside counterthread 23 to create a threaded coupling with the inside thread 14 of the receptacle. Clearly, the closure of lid 2 on receptacle 1 of the container can also be of different design, e.g. with a bayonet coupling. When lid 2 is fully screwed onto receptacle 1, the vial of radiopharmaceutical is held in place between the bottom 11 of receptacle 1 and the underside of lid 2 so that it cannot move. To this end, as illustrated in FIG. 1, lid 2 is hollow on the inside. It presents a cyclindrical upper compartment 24 with a diameter slightly larger than that of vial mouth 30, flaring downwards into a hollow truncated-cone portion 25 that follows the profile of portion 33 of the vial between mouth 30 and cylindrical wall 31. Moreover, as is better illustrated in FIG. 2, lid 2 presents, above its cylindrical upper compartment 24, a central through-hole 26 with a diameter close to that of the central portion of the rubber cap of radiopharmaceutical vial 3 which is accessible for the insertion of an aspiration needle. To facilitate this operation, central through-hole 26 has an outward-facing upper flared portion 27. According to the invention described herein, at least receptacle 1, but preferably also lid 2, is made of transparent material. In this way, an operator can check the contents of the vial of radiopharmaceutical and its volume without having to remove lid 2 and lift up the vial. The dose can therefore be calculated on the basis of the concentration (activity/volume) declared by the manufacturer, thereby avoiding the operator having to expose himself to ionising radiation. If the radiation emitted by the radiopharmaceutical is beta-radiation, the material receptacle 1 is made of is polymethyl methacrylate, known under the trade name of plexiglas. Lid 2 can also be made of the same material. Polymethyl methacrylate has excellent shielding characteristics against radioactive emissions, and particularly against beta-emitting isotopes. In addition, polymethyl methacrylate has a low volumic mass and is thus capable of providing a light-weight, easily manageable container. The container has a thickness, both of the wall of the receptacle and that of the lid, that will depend on the beta-emission energy of the isotope it contains. This thickness will be determined by the expert in the sector, simply on the basis of his general knowledge of the subject. In a different realisation of the invention, the radio-pharmaceutical can also consist of mixed emitters, i.e. isotopes that emit both beta and gamma radiation (including 511 KeV annihilation photons), and also those with mixed emission such as, for example, 131I, and il 177Lu. In the particular case of [18F]FDG, in view of its extensive use in clinical practice, the device is particularly suitable for reducing the exposure of health-care operatives to radiation energy. In this case, both the container and the lid will be made of transparent material, either polymethyl methacrylate or glass, rich in lead or tungsten depending on the gamma emission energy. In this case, the second infusion catheter, too, that conveys the radiopharmaceutical to the patient will be housed in appropriately shielded guides. In this particular case, the container and lid will be made of polymethyl methacrylate containing a certain amount of lead such as to ensure the necessary radiation protection and transparency of the receptacle and lid walls. In this realisation, too, the choice of material and determination of the thicknesses of the receptacle and lid walls are matters which come within the field of expertise of the average technician in the sector. The container according to the present invention affords the advantage of allowing the shipment or transportation of precalibrated and customised radiopharmaceuticals for individual patients. Inside the container the operator can check the volume/quantity desired without having to handle the vial. The above-described container allows infusion of the radiopharmaceutical in a patient or its transfer elsewhere without needing to manipulate the vial. The operator, in fact, can extract the radiopharmaceutical with a syringe while the vial containing it remains housed in the container, which affords effective radioprotection. The invention, however, solves the problem posed of allowing infusion in a patient or transfer elsewhere to another radiopharmaceutical receptacle, without needing to aspirate it from its vial with a syringe, and of accurately checking the volume of radiopharmaceutical infused in the patient or transferred to another receptacle. For this purpose, the invention provides a set for infusion of a radiopharmaceutical in a patient or for its transfer elsewhere from its vial housed in the container. The infusion set described above, combined with the container housing the vial of radiopharmaceutical, constitutes a complete kit for managing the radiopharmaceutical without any manipulation and without the operator having to perform a direct extraction operation. Reference is made to FIGS. 3 and 4, that show part of the set and container 1-2 and the set according to the invention in an infusion operation, respectively. The set contains, in combination with container 1-2 of a vial of radiopharmaceutical 3, a conventional bottle 4 containing saline, an infusion catheter and a second infusion catheter, marked collectively 5 and 6, respectively. The saline bottle 4 may be, for example, 250 ml. Details regarding the use of the saline solution will be provided here below. The first infusion catheter 5 is conventionally equipped with twin connectors, with a first needle 50, a flow regulator 51 and a second needle 52. Needle 50 is of known type, suitable for insertion in the bottle of saline solution 4 and is connected to a drop-counter 53. The drop-counter is connected via a small tube 54, and connector 55, to the second needle 52, which is a metal infusion needle. The second infusion catheter 6, according to the invention described herein, is equipped with twin connectors, with a first needle 60, a flow regulator 61 and a second needle 62. Needle 60 is of the infusion type and is connected via connector 63 and small tube 64 to the second needle 62, which is also an infusion needle, via connector 65. In an infusion operation illustrated in FIG. 4, saline bottle 4 is conventionally suspended in a cradle 7 attached to a stand 8, equipped with a support shelf 9. The first infusion catheter is inserted with the first needle 50 in the cap of bottle 4, while the second needle 52 is inserted, via flared portion 27 and central through-hole 26 of lid 2, into the cap of radiopharmaceutical vial 3 in such a way as not to be immersed in the pharmaceutical. As shown in FIG. 5, which is an enlarged view of a detail of FIG. 4, the initial level of radiopharmaceutical is marked L. The second infusion catheter 6 also has its first needle 60 inserted via flared portion 27 and through-hole 26 of lid 2, into the cap of the vial of radiopharmaceutical, whereas the second needle 62 is inserted in the brachial vein B of a patient. The first needle 60 is long enough to touch the bottom of the vial of radiopharmaceutical, where it must be held in place for the complete extraction of the radiopharmaceutical, as shown in FIG. 5. The provision of flow via the bottle of saline solution 4, the first infusion catheter 5, vial 3 in container 1-2, and the second infusion catheter 6 allows the radiopharmaceutical to be delivered by gravity. The saline solution is fed from bottle 4 into radiopharmaceutical vial 3 with flow regulation by means of flow-regulator 51. The influx of saline brings about an increase in pressure in radiopharmaceutical vial 3 which has its entire contents aspirated by the second infusion catheter 6, the flow rate of which is regulated by flow-regulator 61. If one desires to transfer the radiopharmaceutical elsewhere, the transfer is accomplished using air or some other suitable gaseous liquid as the vector fluid. For this purpose, either the infusion catheter which is part of the present invention or any other suitable means can be used. The same kit described above can be used for the transfer of the radiopharmaceutical from its vial to another receptacle, for example in order to fractionate the doses, using air as the driving medium. Disposal of the kit is also risk-free for the operator. The infusion catheters, and particularly the second infusion catheter, are destined to be treated as hazardous materials, as is the radiopharmaceutical vial. After extracting the catheters and unscrewing the lid, the radiopharmaceutical vial is dropped out of its container into the radioactive waste collector, while the container according to the invention can be reused. In addition, the container according to the invention is suitable for use with automatic and even robotic systems for the preparation of individual doses. The container according to the invention and its infusion set are also suitable for managing generally toxic drugs, such as, for example, anticancer agents. |
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abstract | A system and method for storing multiple canisters containing high level waste below grade that afford adequate ventilation of the spent fuel storage cavity. In one aspect, the invention is a ventilated system for storing high level waste emitting heat, the system comprising: an air-intake shell forming an air-intake cavity; a plurality of storage shells, each storage shell forming a storage cavity; a lid positioned atop each of the storage shells; an outlet vent forming a passageway between an ambient environment and a top portion of each of the storage cavities; and a network of pipes forming hermetically sealed passageways between a bottom portion of the air-intake cavity and at least two different openings at a bottom portion of each of the storage cavities such that blockage of a first one of the openings does not prohibit air from flowing from the air-intake cavity into the storage cavity via a second one of the openings. |
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description | 1. Field of the Invention The present invention relates to a spacer grid for supporting nuclear fuel rods of a nuclear fuel assembly charged into a nuclear reactor and, more particularly, to a spacer grid for more close-spaced nuclear fuel rods than conventional ones, in which the supports of each grid strap are located at positions shifted from a central line of each grid strap toward a center of a sub-channel in order to accommodate a reduced gap between the nuclear fuel rods due to the use of dual-cooled nuclear fuel rods, which have excellent cooling performance. 2. Description of the Prior Art FIG. 1 is a perspective view schematically illustrating a conventional nuclear fuel assembly. FIG. 2 is a cross-sectional view schematically illustrating a conventional nuclear fuel assembly. FIG. 3 is a top plane view schematically illustrating a spacer grid applied to a conventional nuclear fuel assembly. FIG. 4 is a perspective view schematically illustrating a spacer grid applied to a conventional nuclear fuel assembly. FIG. 5 is a perspective view schematically illustrating a unit grid strap for a spacer grid supporting a conventional nuclear fuel assembly. As illustrated in the figures, the conventional nuclear fuel assembly 100 comprises nuclear fuel rods 110, guide tubes 140, spacer grids 150, a top end piece 120, and a bottom end piece 130. Here, each nuclear fuel rod 110 has a cylindrical uranium sintered compact in a clad pipe of zircaloy (zirconium alloy). This uranium sintered compact is fissioned to generate high temperature heat. Meanwhile, each guide tube 140 is used as a passage for a control rod, which moves up and down in order to control the output power of a reactor core and to stop the fission reaction. The spacer grid 150 is one of the components constituting the nuclear fuel assembly in a nuclear reactor, and includes a plurality of unit grid straps, each of which has a spring 118 and dimples 119 and functions to support and protect the nuclear fuel rods 110 so that they are arranged at designated positions. When the spring force of the spring 118 and the dimples 119 is too weak, each nuclear fuel rod 110 cannot be arranged at a designated position, and thus has a possibility of losing sound supporting performance. In contrast, when the spring force of the spring 118 and the dimples 119 is too strong, each nuclear fuel rod 110 undergoes defects such as scratching on the surface of the clad tube due to excessive frictional gripping force when it is inserted into the spacer grid. Further, during the operation of the nuclear reactor, the nuclear fuel rods 110 experience longitudinal growth by means of the irradiation of neutrons. This longitudinal growth is not properly accommodated, and thus the nuclear fuel rods 110 are bent. In this manner, when the nuclear fuel rods 110 are bent, they come nearer to or contact neighboring nuclear fuel rods 110. Thus, the coolant channel between the nuclear fuel rods becomes narrow or is blocked. As a result, the heat generated from the nuclear fuel rods is not effectively transmitted to the coolant, thereby increasing the temperature of the nuclear fuel rods. As such, the possibility of generating Departure from Nucleate Boiling (DNB) is increased, which is mainly responsible for the reduction of nuclear fuel output power. The top end piece 120 and the bottom end piece 130 fixedly support the nuclear fuel assembly 100 on upper and lower structures of the reactor core. The bottom end piece 130 includes a screening device (not shown) for filtering foreign materials floating in the reactor core. Meanwhile, each spacer grid 150 is usually made of zircaloy, and includes nuclear fuel rod cells, which support the nuclear fuel rods 110, and guide tube cells, into which the guide tubes 140 are inserted. Each nuclear fuel rod cell is designed to support each nuclear fuel rod 110 at a total of six supporting points using a total of two grid springs 118, which are located on two respective faces, and a total of four dimples 119, which are located in pairs above and below the two grid springs 118 and on the other two respective faces. A cylindrical uranium dioxide compact is inserted into each nuclear fuel rod 110, and the coolant rapidly flows from the bottom to the top of the reactor core in an axial direction through sub-channels 115, each of which is enclosed by four nuclear fuel rods 110 or by three nuclear fuel rods 110 and one guide tube 140. Here, each sub-channel 115 refers to a space enclosed by the nuclear fuel rods 110, and particularly a passage through which a fluid can freely flow to the neighboring sub-channel because it has an open side. Meanwhile, as illustrated in FIGS. 6 and 7, a dual-cooled nuclear fuel rod 10 having an annular structure instead of the cylindrical nuclear fuel rod 110 is disclosed in U.S. Pat. Nos. 3,928,132 and 6,909,765. Here, the dual-cooled nuclear fuel rod 10 having an annular structure includes a sintered compact 11 having an annular shape, an inner clad tube 12 enclosing the inner circumference of the sintered compact 11, and an outer clad tube 13 enclosing the outer circumference of the sintered compact 11. Thus, the coolant flows outside and inside the dual-cooled nuclear fuel rod 10, so that heat transfer is doubled. As a result, the dual-cooled nuclear fuel rod 10 can maintain a low fuel's centerline temperature, and provide high combustion and high output power. In this manner, in the case in which the centerline temperature of the dual-cooled nuclear fuel rod 10 is maintained low, the possibility of damaging the fuel due to an increase in the core temperature of the nuclear fuel is lowered, so that the safety allowance of the dual-cooled nuclear fuel rod 10 can be increased. However, in order to make the dual-cooled nuclear fuel rods 10 structurally compatible with an existing pressurized water reactor (PWR) core, the gap between the nuclear fuel rods becomes considerably narrower compared to that between existing nuclear fuel rods because the positions of the guide tubes 140 cannot be changed in the nuclear fuel assembly 100, and because the outer diameter of each nuclear fuel rod is increased. For example, in the case in which the nuclear fuel assembly is formed according to a candidate design draft for the dual-cooled nuclear fuel rods having a 12×12 array, the gap between the nuclear fuel rods is reduced from 3.35 mm, which is the size of the existing gap, to about 1.24 mm. Thus, due to the narrow gap between the nuclear fuel rods, the spacer grid that has been developed to date cannot be used as that for the dual-cooled nuclear fuel rods 10. In other words, after the thickness of the unit grid strap of the existing spacer grid, which is 0.475 mm, is subtracted from the gap of 1.24 mm between the nuclear fuel rods, the obtained result is again divided by two. Consequently, the gap between the unit grid strap and the nuclear fuel rod is no more than about 0.383 mm. Thus, it is impossible to apply such a shape and a supporting position as in an existing leaf spring within this narrow gap to design a spring having spring rigidity and hydraulic characteristic (mainly pressure loss), which an existing supporting structure has. Accordingly, the present invention has been made keeping in mind the above problems occurring in the prior art, and an object of the present invention is to provide a spacer grid, capable of sound supporting nuclear fuel rods set in a narrow array, unlike an existing spacer grid, which has a problem in which, as the outer diameter of each nuclear fuel rod increases due to the use of dual-cooled nuclear fuel rods for improving cooling performance and obtaining high combustion and high output power, the gap between neighboring nuclear fuel rods is narrowed to thus make it impossible to use the existing spacer grid. In order to achieve the above object, according to one aspect of the present invention, there is provided a spacer grid for close-spaced nuclear fuel rods, in which a plurality of unit grid straps, each of which has supports for supporting each of the nuclear fuel rods set in a narrow array and has a sheet shape, are combined with each other. The supports are located at positions shifted from the longitudinal central line of each unit grid strap toward sub-channels. Here, each support may have a cantilever leaf spring shape, a semi-spherical or elliptical shape, or a semi-cylindrical shape in the longitudinal direction of the unit grid strap. Further, the supports may be formed at the front and rear of each unit grid strap in symmetry based on the longitudinal central line of each unit grid strap. Also, the cantilever leaf spring may include a protrusion, which protrudes from the unit grid strap toward the dual-cooled nuclear fuel rod, and an extension, which extends from the protrusion so as to enclose the outer circumference of the dual-cooled nuclear fuel rod. Further, the extension may have the same curvature as the outer circumference of the dual-cooled nuclear fuel rod, and the protrusion may have a radius of curvature smaller than that of the extension. Reference will now be made in greater detail to exemplary embodiments of the invention, which are illustrated in the accompanying drawings. Wherever possible, the same reference numerals will be used throughout the drawings and the description to refer to the same or like parts. According to an exemplary embodiment of the present invention, a spacer grid 20 supports dual-cooled nuclear fuel rods 10 having a 3×3 array, which is taken as an example. FIGS. 8 through 10 illustrate a spacer grid according to a first embodiment of the present invention. FIG. 8 is a top plan view illustrating a spacer grid having a cantilever leaf spring support, FIG. 9 is a perspective view illustrating a spacer grid having a cantilever leaf spring support, and FIG. 10 is a perspective view illustrating a unit grid strap having a cantilever leaf spring support. The spacer grid 20 of the present invention is a combination of unit grid straps 21, each of which has supports 23 for supporting a dual-cooled nuclear fuel rod 10 by coming into contact with the outer circumference of the dual-cooled nuclear fuel rod 10, and has a sheet shape. The dual-cooled nuclear fuel rod 10 includes a sintered compact 11 having an annular shape, an inner clad tube 12 enclosing the inner circumference of the sintered compact 11, and an outer clad tube 13 enclosing the outer circumference of the sintered compact 11. In particular, each support 23 is located at a position shifted from the longitudinal central line 22 of each unit grid strap 21 toward a sub-channel 115 so as to support the outer circumference of the dual-cooled nuclear fuel rod 10. Here, the sub-channel 115 refers to a space enclosed by four dual-cooled nuclear fuel rods 10 or by three dual-cooled nuclear fuel rods 10 and one guide tube 140. Coolant flows in the axial direction through the sub-channel 115. Preferably, each support 23 has the shape of a cantilever leaf spring 25, and has a predetermined radius of curvature along the outer circumference of the dual-cooled nuclear fuel rod 10 so as to be able to come into surface contact with the dual-cooled nuclear fuel rod 10. To this end, each support 23 includes a protrusion 26, which protrudes from the unit grid strap 21 toward the dual-cooled nuclear fuel rod 10, and an extension 27, which extends from the protrusion 26 so as to enclose the outer circumference of the dual-cooled nuclear fuel rod 10. In other words, the protrusion 26 and the extension 27 are integrally formed with each other by cutting the unit grid strap 21 in a “C” shape at a position shifted from the longitudinal central line 22 of each unit grid strap 21 toward the center of the sub-channel 115, and by bending the cut piece. At this time, the extension 27 is preferably formed so as to have the same radius of curvature as the outer circumference of the dual-cooled nuclear fuel rod 10 for surface contact with the dual-cooled nuclear fuel rod 10. It is advantageous in view of the rigidity of the spring that the protrusion 26 be formed so as to have an arcuate shape having a predetermined radius of curvature smaller than that of the extension 27. Further, the supports 23 are formed at the front and rear of each unit grid strap 21, respectively. The front support 23 and the rear support 23 are in axial symmetry based on the longitudinal central line 22 of each unit grid strap 21. In detail, the front support 23 is adapted to support the dual-cooled nuclear fuel rod 10, which is located at the front of each unit grid strap 21, and the rear support 23 is adapted to support the dual-cooled nuclear fuel rod 10, which is located at the rear of each unit grid strap 21. Thus, the dual-cooled nuclear fuel rod 10 is supported at a total of four positions by the four unit grid straps 21 enclosing one dual-cooled nuclear fuel rod 10. The supports 23 are formed at the front and rear of each unit grid strap 21 so as to have axial symmetry based on the longitudinal central line 22 of each unit grid strap 21, so that they can avoid structural interference or loss of the function as a spring. When formed by a pressing process, the supports 23 minimize disorientation (or flexure) caused by residual strain and residual stress, and are complementarily deformed. Further, the supports 23, having the shape of the cantilever leaf spring 25 formed at the front and rear of each unit grid strap 21, are preferably spaced apart from each other at a predetermined interval so as to be able to maintain proper rigidity when functioning as the spring supporting the dual-cooled nuclear fuel rod 10. FIGS. 11 through 13 illustrate a spacer grid according to a second embodiment of the present invention. FIG. 11 is a top plan view illustrating a spacer grid having semi-spherical shape supports, FIG. 12 is a perspective view illustrating a spacer grid having semi-spherical supports, and FIG. 13 is a perspective view illustrating a unit grid strap having semi-spherical supports. The shape of each dual-cooled nuclear fuel rod 10 and the position of each support 30 are the same as in the first embodiment of the present invention. However, the shape of each support 30 is different from that of the first embodiment of the present invention. Thus, only the shape of each support 30 will be described in detail. The spacer grid 20 according to a second embodiment of the present invention includes at least one semi-spherical support 30 at a position shifted from the longitudinal central line 22 of each unit grid strap 21 toward the center of each sub-channel 115. The semi-spherical supports 30 are formed in a semi-spherical or elliptical shape, and protrude from the front and rear of each unit grid strap 21. At this time, preferably, among the semi-spherical supports 30 having the semi-spherical or elliptical shape, two are formed at the front of the unit grid strap 21 in a vertical direction, and one is formed at the rear of the unit grid strap 21. Alternatively, the semi-spherical supports 30 protruding from the front and rear of the unit grid strap 21 may be formed such that one thereof is located at each of the front and rear of the unit grid strap 21 so as to be in axial symmetry based on the longitudinal central line 22 of the unit grid strap 21. Thereby, when formed by a pressing process, the semi-spherical supports 30 minimize disorientation caused by residual strain and residual stress, and are complementarily deformed. Thus, the dual-cooled nuclear fuel rod 10 can be supported at a total of four positions by the semi-spherical supports 30, which are formed on the four unit grid straps 21 enclosing one dual-cooled nuclear fuel rod 10. FIGS. 14 through 16 illustrate a spacer grid according to a third embodiment of the present invention. FIG. 14 is a top plan view illustrating a spacer grid having semi-cylindrical supports, FIG. 15 is a perspective view illustrating a spacer grid having semi-cylindrical supports, and FIG. 16 is a perspective view illustrating a unit grid strap having semi-cylindrical supports. The shape of each dual-cooled nuclear fuel rod 10 and the position of each support 31 are the same as in the first and second embodiments of the present invention. However, the shape of each support 31 is different from that of the first embodiment of the present invention. Thus, only the shape of each support 31 will be described in detail. The spacer grid 20 according to a third embodiment of the present invention includes semi-cylindrical supports 31 at positions shifted from the longitudinal central line 22 of each unit grid strap 21 toward the sub-channels 115. The semi-cylindrical supports 30 protrude from the front and rear of each unit grid strap 21. Similarly, the semi-cylindrical supports 31 are preferably formed at the front and rear of the unit grid strap 21 so as to be in axial symmetry based on the longitudinal central line 22 of the unit grid strap 21. Thereby, when formed by a pressing process, the semi-cylindrical supports 31 minimize disorientation caused by residual strain and residual stress, and are complementarily deformed. Thus, the dual-cooled nuclear fuel rod 10 is supported at a total of four positions by the semi-cylindrical supports 31, which protrude from the four unit grid straps 21 enclosing one dual-cooled nuclear fuel rod 10. As described above, the dual-cooled nuclear fuel rods 10, arrayed at narrow intervals, are supported by the spacer grid 20 having the supports 23, which are located at the positions shifted from the longitudinal central line 22 of each unit grid strap 21 toward the sub-channels 115, so that the dual-cooled nuclear fuel rods can be stably supported despite the narrow intervals. Moreover, it is not necessary to change the positions of the guide tubes 140 in the nuclear fuel assembly 100, and the top end piece 120 and the bottom end piece 130 can be used as they are. Thus, the spacer grid can be more compatible with the constituents of an existing nuclear fuel assembly 100. Further, the supports 23 are formed into cantilever leaf springs 25, so that they can come into surface contact with the dual-cooled nuclear fuel rod 10, thereby inhibiting fretting attrition from being generated by vibration of the nuclear fuel rods which is caused by flow of the coolant. Moreover, the present invention can be applied to a piping system for transporting fluids and its supporting structures, and ordinary industrial machinery using boilers or heat exchangers. For example, when elongate rods or pipes are arranged at narrow intervals, the present invention is used as one of the shapes of the supports for supporting these rods and pipes. Although exemplary embodiments of the present invention have been described for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims. |
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claims | 1. A laser beam machining method of irradiating an object to be machined with machining laser light beam, from a specific irradiation direction, said method comprising steps of:irradiating the surface of the object with a plurality of adjustment laser light beams as reference light beams for adjusting a converging point of the machining laser light beam, from respective irradiation directions different from each other and further through optical paths different from an optical path of the machining laser light beam, the plurality of adjustment laser light beams having a relationship such that a beam spacing therebetween on the surface of the object changes according to a depth position of the converging point from the surface of the object, irradiation areas of the plurality of adjustment laser light beams satisfying a specific positional relationship with respect to the converging point of the machining laser light beam;monitoring irradiation areas of the plurality of adjustment laser light beams, on the surface of the object;exacting a positional information of the surface of the object with respect to the converging point of the machining laser light beam, on the basis of information of the irradiation areas and the specific positional relationship;moving the object so as to make the beam spacing obtained from the positional information become a predetermined beam spacing; andirradiating the object, whose position has been adjusted, with the machining laser light beam, after moving the object. 2. The laser beam machining method according to claim 1, wherein the plurality of the adjustment laser light beams are applied so as to intersect at the converging point of the machining laser light beam. 3. The laser beam machining method according to claim 1, wherein a laser irradiation condition of the machining laser light beam is changed in accordance with the depth position of the converging point of the machining laser light beam from the surface of the object. 4. A laser beam machining apparatus of executing a method according to claim 1, said apparatus comprising:a machining laser light source for outputting machining laser light beam for irradiating the object;a converging optical system for converging the machining laser light beam outputted from said machining laser light source;an adjustment laser light source for outputting a plurality of adjustment laser light beams as reference light beams for adjusting a converging point of the machining laser light beam;an adjustment optical system for guiding the plurality of adjustment laser light beams from respective directions different from each other to the object, while irradiation areas of the plurality of adjustment laser light beam satisfy a specific positional relationship with respect to the converging point of the machining laser light beam;an adjustment section for moving the object;a measurement section for monitoring respective irradiation areas of the plurality of adjustment laser light beams on the object; anda control section for exacting a positional information of the surface of the object with respect to the converging point of the machining laser light beam, on the basis of information of the irradiation areas and the specific positional relationship, and calculating an amount of control for controlling said adjustment section so as to make the position of the object with respect to the converging point of the machining laser light beam correspond to a preset position. 5. The laser beam machining apparatus according to claim 4, wherein said adjustment section guides the plurality of adjustment laser light beams so as to intersect with each other at the converging point of the machining laser light beam. 6. The laser beam machining apparatus according to claim 4, wherein said adjustment section includes a drivers stage on which the object is mounted. 7. The laser beam machining apparatus according to claim 4, wherein said control section calculates a direction and amount of inclination of the object, on the basis of respective deformation ratios of the measured irradiation area forms of the plurality of adjustment laser light beams with respect to cross-sectional beam forms thereof. 8. The laser beam machining apparatus according to claim 4, wherein a laser irradiation condition of the machining laser light beam is changed in accordance with the depth position of the converging point of the machining laser light beam from the surface of the object. 9. The laser beam machining method according to claim 1, wherein inclinatory degree and inclinatory direction of the object are adjusted such that the beam cross-section of the one adjustment laser light beam becomes a preset target shape. 10. A laser beam machining method of irradiating an object to be machined with machining laser light beam, from a specific irradiation direction, said method comprising steps of:irradiating the surface of the object with a plurality of adjustment laser light beams as reference light beams for adjusting a converging point of the machining laser light beam, from respective irradiation directions different from each other and further through optical paths different from an optical path of the machining laser light beam, while irradiation areas of the plurality of adjustment laser light beams satisfy a specific positional relationship with respect to the converging point of the machining laser light beam;monitoring irradiation areas of the plurality of adjustment laser light beams, on the surface of the object;exacting a positional information of the surface of the object with respect to the converging point of the machining laser light beam, on the basis of information of the irradiation areas and the specific positional relationship;moving the object so as to adjust the state of installation of the object with respect to the converging point of the machining laser light beam, on the basis of the calculated direction and amount of inclination of the object; andirradiating the object, whose position has been adjusted, with the machining laser light beam, after moving the object,wherein at least one of the plurality of the adjustment laser light beams has a beam cross-section with a predetermined shape, andwherein, after exacting information concerning the beam cross-section of the one adjustment laser light beam on the irradiation area by monitoring the associated one of the irradiation areas, an inclinatory condition of the object is adjusted such that the extracted beam cross-section on the irradiation becomes similar to the predetermined shape. 11. A laser beam machining apparatus for executing the method according to claim 10, said apparatus comprising:a machining laser light source for outputting machining laser light beam for irradiating the object;a converging optical system for converging the machining laser light beam outputted from said machining laser light source;an adjustment laser light source for outputting a plurality of adjustment laser light beams as reference light beams for adjusting a converging point of the machining laser light beam;an adjustment optical system for guiding the plurality of adjustment laser light beams from respective directions different from each other to the object, while irradiation areas of the plurality of adjustment laser light beam satisfy a specific positional relationship with respect to the converging point of the machining laser light beam;an adjustment section for moving the object;a measurement section for monitoring respective irradiation areas of the plurality of adjustment laser light beams on the object; anda control section for exacting a positional information of the surface of the object with respect to the converging point of the machining laser light beam, on the basis of information of the irradiation areas and the specific positional relationship, and calculating an amount of control for controlling said adjustment section so as to make the position of the object with respect to the converging point of the machining laser light beam correspond to a preset position. 12. The laser beam machining apparatus according to claim 11, wherein inclinatory degree and inclinatory direction of the object are adjusted such that the beam cross-section of the one adjustment laser light beam becomes a preset target shape. |
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042082471 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention: This invention relates to nuclear reactors, and more particularly to neutron emitting sources utilized for monitoring reactor neutron flux and for core startup purposes. 2. Description of the Prior Art: In order to start up a nuclear core, a neutron source or emitter is typically utilized. The sources used have included those referred to as "primary sources" and "secondary sources". Primary sources are those which are made of neutron emitting isotopes in the form in which they are initially placed in the reactor core. Secondary sources are those which are made of initially non-neutron emitting materials, which become neutron emitters during operation of the reactor core. Typical of the materials utilized for sources are those including combinations of polonium and beryllium, plutonium and beryllium, antimony and beryllium, americium-beryllium and curium, and sources including californium. These sources can be relatively expensive. Because secondary sources are irradiated in-core, they are typically less expensive than the primary sources. In addition to performing a start-up function, the sources are utilized during core shutdown to ensure the operability of monitoring and detection apparatus, such as neutron detectors aligned with the reactor core and positioned outside of the reactor vessel. This is in accordance with governmental regulations applicable to the nuclear industry which dictate that means must be provided for monitoring or otherwise measuring and maintaining control of the fission process under all operating conditions, including shutdown. Accordingly, neutron sources for commercial reactors have been positioned within the nuclear core, and remain within the core, during at least one entire operating cycle. The sources maintain a fixed position. Small test reactors have also utilized sources driven through a hollow shielded tunnel up into the test core, which are subsequently removed at power operation. In the larger reactors, sources are inserted in selected fuel assemblies and extend within fuel assembly guide thimbles designed to receive control elements. They are therefore inserted within fuel assemblies positioned so as not to receive a control element. They are also disposed in assemblies close to the core periphery so as to be positioned close enough to activate the detection and monitoring apparatus outside of the reactor vessel. As the sources remain within an assembly for an entire core cycle, the primary sources, excluding those of californium, burn out within several fuel cycles when exposed to the high neutron flux during power operation. Californium sources have experienced longer life. In other terms, the sources are consumed by neutron induced fission and transmutation, as well as being mechanically damaged by heat, when exposed to the intense neutron flux levels characteristic of power operation. As replacement of primary sources is costly, the secondary sources activated under the high neutron flux are used as replacements. The procedure, however, is costly not only in terms of initial source cost, but also in terms of the time and effort expended to reshuffle any type of source during maintenance or refueling operations. There is also a risk of mechanical damage of a secondary source during handling at refueling, and resulting extended down time. It is therefore desirable to provide neutron sources for nuclear reactors which alleviate costs and time associated with replacement. The sources should desirably have an extended useful life, require minimum reshuffling, and provide adequate neutron emissions for both start-up and monitoring functions. SUMMARY OF THE INVENTION This invention provides an improved neutron source for a thermal nuclear reactor and also an arrangement for utilizing a neutron source in a thermal nuclear reactor, each of which provides an extended useful life. In a preferred embodiment the source includes a radioactive fast neutron emitting material, preferably a combination of plutonium-238 and beryllium, encapsulated in a structurally sound alloy which is essentially black to thermal neutrons, but not to neutrons in other energy ranges. Alloys including cadmium are preferred. As plutonium-238 has a relatively long high-life, approximately 89 years, if protected from the high neutron flux within a reactor it can be utilized as an alpha particle emitter for substantially the life of the plant. The essentially black alloy does protect the plutonium-238 from the high thermal neutron flux. Alpha particles emitted spontaneously by the plutonium-238 react with beryllium nuclei which undergo (.alpha., n) reactions, thereby emitting fast neutrons which pass through the cadmium alloy cladding. The neutrons activate the exvessel neutron detectors. Another embodiment of the invention also includes a unique arrangement providing movable positioning of the inventive, or any, neutron emitting source. A source is affixed to a reciprocating control element which passes into and out of the core, and is particularly affixed to a control element which is raised above the core during power operation and which remains in the core at shutdown. Therefore, during periods of high neutron flux, at power, the sources remain above the core elevation and the area of high flux. The sources can be incorporated in portions of selected control elements or rods which maintain the source at the higher position at the times of high flux, such as the control elements known in the art as "first bank" or "shutdown" rods. At shutdown, the sources are alignable with the detection apparatus which is external to the vessel and aligned with the core, so as to cintinually activate the detection apparatus. |
046648682 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT An embodiment of the present invention will be described below in conjunction with FIGS. 7 and 8, wherein reference numeral 1 denotes a toroidal coil. A bridge portion R off the straight portion of the coil is elongated in the radial direction, and coil support frames 5a, 5b containing the toroidal coil 1 also have shapes along the portion R that is off the straight portion of the toroidal coil 1. Reference numeral 11 denotes a reinforcing member that is welded to wedge-like coupling portions 5c of the coil support frames 5a, 5b, which is formed by grinding a thick plate, or by forging. The reinforcing member 11 is located at the portion R off the straight portions of coil support frames 5a, 5b that is, the portions of the frames bridging the opposite ends of the straight portions to the curved portions thereof; and has such a structure that the wedge-like coupling portions 5c are stretched in the radial direction. The reinforcing member 11 further has opposing keyways 11a, 11b in the oppositing surfaces of adjacent reinforcing members; for coupling the coil support frames 5a, 5b, and spaces 11c behind wall portions of the reinforcing members for fastening the coil support frames 5a, 5b. The keyway 11a is wide, the keyway 11b is narrow, and the pair of these keyways constitute a stepped keyway. Reference numeral 12 denotes a key inserted in the key grooves 11a, 11b to couple the coil support frames 5a, 5b together, and is formed in a stepped shaped. Reference numeral 13 denotes bolts which are screwed into the wall portions of reinforcing member 11 and into the spaces 11c to rigidly fasten the coil support frames 5a, 5b together. The operation of the toroidal coil apparatus according to the present invention will be described herebelow. A plurality of toroidal coils 1 are positioned in a radial manner directly on the rack 7, the wedge surfaces of wedge-like coupling portions 5c of coil support frames 5a, 5b are intimately wedged together, and a force is applied from the back side of the coils by hydraulic jacks, such that the centripetal force is correctly received via the wedge surfaces. Under this condition, the key 12 is inserted in the keyways 11a, 11b to couple the neighboring coil support frames 5a, 5b together. This eliminates slippage between the coil support frames 5a, 5b. The neighboring coil support frames 5a, 5b are then rigidly secured by bolts 13. The above-mentioned operation is effected successively in the torus direction to firmly support the toroidal coils 1. With the thus constructed toroidal apparatus, the wedgelike coupling portions 5c of the coil support frames 5a, 5b establish a construction that extends in the radial direction due to the reinforcing members 11. Further, owing to the bonding force and fastening force by the keys 12 and bolts 13, the straight portions of coils are more firmly secured than those of the conventional art and, further, the same effect is obtained as when inversion preventing beams are provided in the vicinity of portions off the straight portion of the coil. Accordingly, the inverting force F.sub.Q can be strikingly reduced. Since the inverting force F.sub.Q is reduced, the stress can also be reduced in the vicinities of straight portions of the coils. Although the above embodiment has described the case in which the reinforcing members 11 are arranged between the coil support frames 5a, the reinforcing members 11 can also be arranged in the upper portion or the lower portion of the coil support frames 5a to obtain the same effects. According to the present invention as described in the foregoing, reinforcing members having keyways and spaces are provided at portions off the straight portions of the coil support frames to couple and fasten the coil support frames, wherein keys are inserted in the keyways to couple the coil support frames, and the coil support frames are rigidly secured by bolts. It is therefore possible to decrease the stress that results from the inverting force. |
abstract | A micro-manipulation method enables micro-objects to be handled as desired with excellent repeatability when using microscopes that radiate electron beams and use secondary, reflected and transmitted electron signals to magnify objects for observation. For micro-manipulation of micro-objects with a micro-handling tool under electron beam irradiation of an electron microscope, there is a means to adjust the accelerating voltage of the electron beam, the potential of the micro-handling tool and the potential of the work substrate, thereby enabling the effective pick up and release of micro-objects with the micro-handling tool. |
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047643366 | claims | 1. A de-activation pool of a nuclear station, whose walls are provided with a fluid-tight covering of exterior metal sheets welded together by their edges by means of weld beads and supported under each weld bead by a beam in the region of a planar bearing surface of the beam, each beam being fixed in a corresponding wall of the pool; each beam being hollow and, each beam comprising a first section element having roughly the shape of a U in cross-section, the planar bearing surface extending longitudinally in facing relation to at least a part of the bottom of the first section element; and a member disposed inside the first section element and defining, with said first section element, said planar bearing surface and at least two longitudinal channels in the inner volume of the beam, said member having a first segment and a second substantially horizontal segment vertically spaced from said bottom of said first section element and forming part of said planar bearing surface, only one of said channels forming a weld-inspection channel, and at least one of said channels having an open cross-section and forming a draining channel, said channels being located so that said planar being surface is facing and opposite to said channels inside the beam. 2. A pool according to claim 1, wherein the U has two branches and said bottom, and wherein said planar bearing surface and the edges of the branches of the first section element are contained in a common plane. 3. A pool according to cliam 2, wherein the edge of the branches of the section elemtent each has a flange extending in a direction parallel to the bearing surface, outwardly of the beam, said flanges completing the bearing surface by their surface oriented toward the exterior of the beam. 4. A pool according to claim 2, wherein the bearing surface is spaced from each of the branches of the section element. 5. A de-activation pool of a nuclear station, whose walls are provided with a fluid-tight covering of exterior metal sheets welded together by their edges by means of weld beads and supported under each weld bead by a beam in the region of a planar bearing surface of the beam, each beam being fixed in a corresponding wall of the pool and comprising a first section element having roughly the shape of a U in cross-section, the planar bearing surface extending longitudinally in facing relation to at least a part of the bottom of the first section element, and a channel having an open cross-section, characterized in that each beam comprises a member disposed inside the first section element, said member defining, with said first section element, said planar bearing surface and at least two longitudinal channels in the inner volume of the beam; wherein the U has two branches and said bottom, and wherein said planar bearing surface and the edges of the branches of the first section element are contained in a common plane; wherein the bearing surface is spaced from each of the branches of the section element; and wherein each beam comprises an inspection channel having a closed cross section, and two longitudinal draining channels having an open cross-section. wherein the U has two branches and said bottom, and wherein said planar bearing surface and the edges of the branches of the first section element are contained in a common plane; wherein the bearing surface is spaced from each of the branches of the section element; wherein each beam comprises two longitudinal channels having an open cross-section respectively forming an inspection and draining channel and a draining channel; and wherein the member disposed inside the first section element is a second section element which has, in cross-section, roughly the shape of an L having a first branch, fixed in a position perpendicular to the bottom of the first section element, and another branch whose surface faces in a direction away from said bottom and forms said bearing surface. 6. A pool according to claim 5, wherein the member disposed inside said first section element is a second section element which has in cross-section roughly the shape of a U having branches fixed to the bottom of the first section element and having a bottom, facing in a direction opposed to that of the bottom of said first section element, forming said bearing surface. 7. A pool according to claim 6, wherein the branches of the second section element each define a flange so that said second section element has, in cross-section, roughly the shape of an omega. 8. A pool according to claim 4, wherein each beam comprises two longitudinal channels having an open cross-section respectively forming an inspection and draining channel and a draining channel. 9. A de-activation pool of a nuclear station, whose walls are provided with a fluid-tight covering of exterior metal sheets welded together by their edges by means of weld beads and supported under each weld bead by a beam in the region of a planar bearing surface of the beam, each beam being fixed in a corresponding wall of the pool and comprising a first section element having roughly the shape of a U in cross-section, the planar bearing surface extending longitudinally in facing relation to at least a part of the bottom of the first section element, and a channel having an open cross-section, characterized in that each beam comprises a member disposed inside the first section element, said member defining, with said first section element, said planar bearing surface and at least two longitudinal channels in the inner volume of the beam; 10. A pool according to claim 9, wherein the first branch (24) of the second section element (l7a) is fixed to the bottom (19) of the first section element and has a flange (23a) which extends, in a direction opposed to that of said another branch (25), along the bottom (19) of the first section element. 11. A reservoir according to claim 9, wherein each beam is made from a single member by a cold drawing operation. |
summary | ||
description | 1. Field of the Invention The present invention relates to a method for ablating hyaluronan-based hydrogels with X-rays and a method for fabricating three-dimensional microchannels of hyaluronan hydrogels with the X-ray ablation technique. 2. Background of the Related Art Phase transition triggered by external perturbation is quite important for “intelligent materials” and a key issue in diverse fields ranging from biomedicine to chemistry, physics, and materials science. Hydrogels-three-dimensional networks of crosslinked polymer chains-exhibit transitions in response to perturbations such as electric fields, temperature changes, pH changes, concentration changes, enzymes, electron beams, sound, and light. Hydrogels are actively studied with the objective to develop new technologies to control fluidity, viscoelasticity, solvent volatility, optical transmission, ion transport, and other properties. Hyaluronan (salt form of hyaluronic acid, HA) is an important extracellular and cell-surface associated polysaccharide. It is commonly synthesized as a large, negatively charged, linear polysaccharide that is composed of repeating disaccharide units of glucuronic acid and N-acetylglucosamine: [−β(1,4)−GlcUA−β(1,3)−GlcNAc−]n. HA has an important role in tissue homeostasis and biomechanical integrity via remarkable physicochemical characteristics such as viscoelastic and hygroscopic properties. These properties of HA lead to its widespread applications for bioengineered tissue scaffolds. Related physiological functions stimulate interest on its role in cell biology, pathology, immunology, and cancer research. The microfabrication of cell-laden HA architecture to resemble three-dimensional (3D) cellular microenvironments is also an important issue in HA. The HA molecular weight (MW) is, in general term, of critical importance because of its remarkable effects on cell activities. Although low MW HA, required for safe biomedical applications, is produced by enzymatic or non-enzymatic degradations, there are few reports on the safe, effective methods to fabricate 3D architectures of HA hydrogels. As mentioned above, Hyaluronan hydrogels are promising materials for tissue scaffolds or cellular microenvironments, but it is still a great challenge to fabricate three-dimensional architectures. Here we describe a versatile and robust protocol to fabricate three-dimensional microchannels of hyaluronan hydrogels with a finely tunable X-ray ablation technique. The principle of X-ray ablation is that polymer chains rapidly degrade by X-ray irradiation. This protocol will open new opportunities for tunable three-dimensional hydrogel architectures. Therefore, it is the first object of the present invention to provide a method for ablating hyaluronan-based hydrogels with X-rays. And, it is the second object of the present invention to provide a method for fabricating three-dimensional microchannels of hyaluronan hydrogels with X-ray ablation. To accomplish the first object, according to one aspect of the present invention, there is provided a method for ablating hyaluronan-based hydrogels, the method comprising the steps of: (a) preparing hyaluronan-based hydrogels; and (b) performing X-ray irradiation to the hyaluronan-based hydrogels to induce a degradation of the hyaluronan-based hydrogels by a gel-to-sol transition during the X-ray irradiation. Preferably, the X-ray irradiation may be performed using hard X-rays. Preferably, the X-ray irradiation may be performed using X-rays in the range of 10-60 keV. Preferably, the degradation kinetics of the hyaluronan-based hydrogels may be determined by total X-ray dose during the X-ray irradiation. Preferably, the total X-ray dose to initiate the transition may be in the range of 0.2˜1 J g−1. Preferably, the total X-ray dose to complete the transition may be in the range of 2˜4 J g−1. To accomplish the second object, according to another aspect of the present invention, there is provided a method for fabricating three-dimensional microchannels of hyaluronan-based hydrogels with X-ray ablation, the method comprising the steps of: (a) preparing hyaluronan-based hydrogels; and (b) performing X-ray irradiation to the hyaluronan-based hydrogels via a mask transmitting X-rays locally to induce a degradation of the hyaluronan-based hydrogels by a gel-to-sol transition during the X-ray irradiation. Preferably, the X-ray irradiation may be performed using hard X-rays. Preferably, the X-ray irradiation may be performed using X-rays in the range of 10-60 keV. Preferably, the depth and the width of the microchannels may be tunable by adjusting the X-ray dose and the mask width, respectively. Preferably, the degradation kinetics of the hyaluronan-based hydrogels may be determined by the total X-ray dose during the X-ray irradiation. Preferably, the total X-ray dose to initiate the transition may be in the range of 0.2˜1 J g−1. Preferably, the total X-ray dose to complete the transition may be in the range of 2˜4 J g−1. The preferred embodiments of the invention will be hereafter described in detail, with reference to the accompanying drawings. Here, we report a novel protocol using a short X-ray irradiation to ablate bulky HA hydrogels based on well controlled degradation kinetics: ca. 1 mg of HA rapidly degrades within 30 s of hard-X-ray irradiation, with the same specific cleavage as in enzymatic degradation. Based on using such a fast X-ray ablation process, we were able to fabricate three-dimensional HA hydrogel microchannels, as illustrated in FIG. 1. We note that the depth and the width in a single channel are tunable by adjusting the X-ray dose and the mask width, respectively. The X-ray irradiation of HA raw materials and HA-based hydrogels (crosslinked with divinyl sulfone) was performed using synchrotron hard X-rays (10-60 keV), which were also used to image the induced degradation in real time. The fast degradation kinetics is due to a rapid chain scission associated with the formation of carbonyl or carboxyl groups in the HA backbone. In general, the X-ray ablation of the HA-based hydrogels could be quite effective in cleaving bulky HA architecture for 3D cellular microenvironments. FIG. 2a shows schematic views of representative in-situ microradiographs that demonstrate a real time gel-to-sol transition of HA hydrogel crosslinked with divinyl sulfone (DVS) during X-ray irradiation. Spherical silica balls (˜120 μm in diameter) that initially stayed in the top region of the HA hydrogel, fell down to the bottom with irradiation time, clearly indicating the X-ray-induced gel-to-sol transition. The irradiated mass of the HA-DVS hydrogel (MW=232 kDa, density≈1 g cm−3) was ca. 1 mg in a capillary tube (ca. 1200 μm in diameter). As already mentioned, the transition was very fast: this mass degraded within 30 s of irradiation. The ablation kinetics depends on the X-ray dose rate (or flux) and on the irradiation time, as illustrated in the sol-gel phase diagram of FIG. 2b (the X-ray dose rate of FIG. 2a was ca. 1 kGy s−1). However, it is the total X-ray dose (=the X-ray dose rate times the irradiation time) that actually determines the kinetics. The total dose required to ‘initiate’ the transition is approximately fixed and ≈0.2˜1 J g−1 (=kGy) (from the lower dashed line), as shown in FIG. 2b. The total dose to ‘complete’ the transition is also approximately fixed, ≈2˜4 J g−1 (from the upper dashed line). This suggests that the total energy (total X-ray dose) required for the transition is constant even if the flux changes. At a total dose of 1˜2 J g−1 the gel and the sol states coexist. These findings are important since they show that the degradation kinetics can be controlled by modifying the total dose. We explain the principle of the X-ray ablation process. The HA-DVS hydrogel degradation at ca. 1 kGy s−1 was monitored with UV and FT-IR spectroscopes (FIG. 3). The absorption band at 260˜270 nm (indicated by the arrow) in the UV spectra is due to carbonyl or carboxyl groups. The intensity increase of the band with the irradiation time is due to the increase of the total dose-(see FIG. 2b). As shown in the FT-IR spectra in the inset of FIG. 3, a similar increase with the irradiation time is observed for the absorption band at 1700˜1750 cm−1 (indicated by the gray zone) that also corresponds to carbonyl or carboxyl groups. The same band evolutions were found in the UV and FT-IR spectra of the HA raw material (powders or solutions, MW=232 kDa) (FIG. 4). The UV and the FT-IR spectra of the two HA and HA-DVS hydrogel samples suggest that the X-ray irradiation cleaves the HA backbone. We note that the band evolution within one minute is quite marked in the UV and the FT-IR spectra, indicating that the irradiation-induced chain scission is very rapid. As shown in FIG. 5, gel permeation chromatography (GPC) detected a significant reduction of the molecular weight by X-ray irradiation in the HA and the HA DVS hydrogel samples. The splitting of the GPC spectra for hard X-rays is similar to the depolymerization process of HA by soft X-rays. This result corroborates the conclusion that the HA backbone is cleaved by X-ray irradiation. Also note that the X-ray-induced degradation of the HA-DVS hydrogels corresponds to no significant changes in the FT-IR spectra except for the band at 1700˜1750 cm−1, similar to enzyme-induced HA degradation. This indicates that the X-ray ablation process results from the controlled degradation of specific chains in the HA molecules. The formation of carbonyl or carboxyl groups is attributed to the scission of glycosidic linkages between monosaccharide units in HA. We thus conclude that the fast degradation kinetics is due to a rapid chain scission associated with the formation of carbonyl or carboxyl groups in the HA backbone. To summarize, we presented a novel protocol for microfabrication of HA-based hydrogels with a short hard-X-ray irradiation (X-ray ablation). This protocol could be quite effective in cleaving bulky HA architecture for 3D cellular microenvironments. Compared to other approaches such as laser ablation, X-ray irradiation offers the advantages of high penetration, local irradiation, non-thermal character, and remote control—possibly opening up new opportunities in 3D HA hydrogel microfabrication for a variety of biological and medical applications. In the preferred embodiments of the present invention are used experimental conditions as follows: Materials: Sodium hyaluronate, sodium salt of hyaluronic acid (HA), with a molecular weight (MW) of 234 kDa was purchased from Lifecore Co. (Chaska, Minn.). HA with a MW over 2 million, under the trade name of Suvenyl®, was obtained from Chugai Pharmaceutical Co. (Tokyo, Japan). Divinyl sulfone (DVS) was purchased from Tokyo Chemical Industry Co. (Tokyo, Japan). Sodium hydroxide (NaOH) and methanol were obtained from Wako Pure Chemical Industries (Osaka, Japan). All the chemicals were used without further purification. HA-DVS hydrogel preparation: HA (68 mg) was dissolved in 1.68 mL of 0.2N NaOH (pH=13). After complete dissolution, 20.02 μL of DVS was added to the HA solution for the crosslinking reaction with the hydroxyl groups of HA. The molar ratio of DVS to hydroxyl group was 1:1. The final precursor solution was mixed completely, and 100 μL of the solution were inserted into each one of 15 syringes (volume=1 mL). After incubation at 37° C. for 1 h to complete the crosslinking reaction for HA-DVS hydrogel preparation, the syringes were sealed with prewashed dialysis membrane tube (MWCO of 7 kDa) and dialyzed against PBS for 24 h. The ions (Na+ and OH−) diffused out through the dialysis membrane neutralizing the pH inside HA-DVS hydrogels X-ray irradiation and real-time phase-contrast X-ray microscopy: The X-ray irradiation and the real-time phase-contrast X-ray imaging were performed using hard X-rays (10-60 keV) at the 7B2 beamline available at the Pohang Light Source (PLS) 2.5 GeV, 150 mA storage ring in Pohang, Korea. Spatially-coherent synchrotron X-rays were used to track the detailed gel-to-sol transition during X-ray irradiation, using a CdWO4 scintillator crystal and a CCD (charge-coupled device) camera. The scintillator-specimen distance was set at 150 mm to optimize phase-contrast enhancement. The beam spot size was tuned to 1.50×1.13 mm2 and the microradiology spatial resolution was 0.5 μm. The X-ray dose rate was controlled by adding silicon attenuators and measured with a previously calibrated ion chamber. The sequential microradiographs were taken with an interval time (acquisition time of 0.1 s and data transmission time of 0.4 s) of 0.5 s. Sequential snapshots in a movie were treated with the Image-Pro Plus software. UV, FT-IR, GPC measurements: UV absorption spectra were obtained using SHIMADZU UV-2550 spectrophotometer at the range of 220-600 nm. FT-IR spectra were measured at a spectral resolution of 4 cm−1 with a Bomem DA8 FT-IR spectrometer equipped with a liquid nitrogen-cooled MCT detector. GPC analysis was performed using the following system: Waters 1525 binary HPLC pump, Waters 2487 dual λ absorbance detector, Waters 717 plus autosampler, Ultrahydrogel TM 1000 and TM 250 columns (7.8 mm×30 cm) (Milford, Mass., USA). Eluant was 34 mM phosphate buffer (pH 6.6)/methanol=80:20 (v/v) and the flow rate was 1 mL/min. Detection wavelength was 210 nm. While the present invention has been described with reference to the particular illustrative embodiments, it is not to be restricted by the embodiments but only by the appended claims. It is to be appreciated that those skilled in the art can change or modify the embodiments without departing from the scope and spirit of the present invention. |
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claims | 1. An absorber cluster for a pressurized water nuclear reactor, comprising:a cluster of absorber rods, each absorber rod comprising a tubular sheath and an absorber column formed by a stack of neutron absorbing elements received within the tubular sheath, each of the neutron absorbing elements extending along a longitudinal axis and having a cross-sectional structure determined by a cross-sectional geometry, a composition and a porosity of the neutron absorbing element, the absorber columns containing at least two absorber assemblies and each of the absorber assemblies consists of a plurality of the neutron absorbing elements having a same cross-sectional structure, the absorber rods being configured for being movable vertically during normal operation of the pressurized water nuclear reactor in order adjust a reactivity of the core of the pressurized water nuclear reactor,the cluster comprising at least a first absorber assembly of the absorber assemblies and a second absorber assembly of the absorber assemblies, the neutron absorbing elements respectively of the first absorber assembly and the second absorber assembly being formed from a same material or a same combination of materials selected from the group of neutron absorbing materials consisting of a first europium hafnate, a second europium hafnate, a first samarium hafnate, a second samarium hafnate, hafnium carbide and samarium hexaboride,the first europium hafnate and the second europium hafnate being two defined compounds each formed by a reaction between hafnium oxide and europium oxide, the first europium hafnate and the second europium hafnate having different compositions,the first samarium hafnate and the second samarium hafnate being two defined compounds each formed by a reaction between hafnium oxide and samarium oxide, the first samarium hafnate and the second samarium hafnate having different compositions,the cross-sectional structure of the neutron absorbing elements of the first absorber assembly being different from the cross-sectional structure of the neutron absorbing elements of the second absorber assembly before a first use of the absorber cluster in the pressurized water nuclear reactor. 2. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the first absorber assembly and the second absorber assembly are formed from the first samarium hafnate, and the first samarium hafnate has a molar content of samarium oxide of between 8% and 20%. 3. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the first absorber assembly and the second absorber assembly are formed from the first europium hafnate, and the first europium hafnate has a molar content of europium oxide of between 43% and 54%. 4. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the absorber assemblies are made from the first europium hafnate and/or second hafnate europium. 5. The absorber cluster according to claim 1, wherein the neutron absorbing elements of the first absorber assembly are made exclusively of neutron absorbing materials of the group consisting of the first samarium hafnate, the second samarium hafnate, the first europium hafnate and second europium hafnate, while the neutron absorbing elements of the second absorber assembly are made exclusively of hafnium carbide. 6. The absorber cluster according to claim 5, wherein the neutron absorbing elements of the first absorber assembly and the neutron absorbing elements of the second absorber assembly are disposed in distinct absorber rods. 7. The absorber cluster according to claim 6, wherein the absorber rods include as many absorber rods receiving neutron absorbing elements of the first absorber assembly as absorber rods receiving the neutron absorbing elements of the second absorber assembly. 8. The absorber cluster according to claim 5, wherein the neutron absorbing elements of the first absorber assembly and the neutron absorbing elements of the second absorber assembly are arranged in distinct absorber rods which are adjacent in pairs. 9. The absorber cluster according to claim 5, wherein the absorber assemblies of the absorber cluster are exclusively the first absorber assembly and the second absorber assembly. 10. The absorber cluster according to claim 5, wherein the absorber assemblies include a third absorber assembly comprising a plurality of the neutron absorbing elements made exclusively of boron carbide, exclusively of samarium hexaboride, or exclusively of materials of the group consisting of the first europium hafnate and the second europium hafnate. 11. The absorber cluster according to claim 10, wherein the neutron absorbing elements of the third absorber assembly are arranged in the absorber rods receiving the neutron absorbing elements of the first absorber assembly and the second absorber assembly. 12. The absorber cluster according to claim 10, wherein the neutron absorbing elements of the third absorber assembly are arranged in the absorber rods receiving the neutron absorbing elements of the first absorber assembly and the second absorber assembly in the upper part of the absorber rods. 13. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the first absorber assembly and the second absorber assembly contain a plurality of different materials from among the group of materials. 14. The absorber cluster according to claim 1, wherein at least one of the absorber columns exclusively contains neutron absorbing elements of a same absorber assembly. 15. The absorber cluster according to claim 1, wherein the neutron absorbing elements of the first absorber assembly and the second absorber assembly are made from the same material or the same combination of materials of the group of materials, and differ in the geometry of their cross-section and/or porosity rate. 16. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the absorber assemblies contain exclusively materials of the group consisting of the first europium hafnate and the second europium hafnate. 17. The absorber cluster according to claim 1, wherein at least one of the absorber assemblies has a plurality of the neutron absorbing elements having a multilayer cross-sectional structure comprising a plurality of layers each made of a material or a combination of materials from among the group of materials, wherein each of the neutron absorbing element having a multilayer cross-sectional structure comprises at least two layers of distinct compositions and/or distinct porosity rates. 18. The absorber cluster according to claim 1, wherein at least one of the absorber columns contains the neutron absorbing elements of the first absorber assembly and the neutron absorbing elements of the second absorber assembly that are distributed along the absorber column in a repeated pattern. 19. The absorber cluster according to claim 1, wherein the first europium hafnate, the second europium hafnate, the first samarium hafnate, the second samarium hafnate, the hafnium carbide and/or the samarium hexaboride contained in the neutron absorbing elements of the first absorber assembly and the second absorber assembly are in cubic crystalline phase. 20. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the first absorber assembly and the second absorber assembly contain a single material of the group of materials. 21. The absorber cluster according to claim 1, wherein at least one absorber columns contains neutron absorbing elements of at least two distinct absorber assemblies. 22. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the first absorber assembly and the second absorber assembly are formed from the second samarium hafnate, and the second samarium hafnate has a molar content of samarium oxide of between 50% and 60%. 23. The absorber cluster according to claim 1, wherein the neutron absorbing elements of at least one of the first absorber assembly and the second absorber assembly are formed from the second europium hafnate, and the second europium hafnate has a molar content of europium oxide of between 19% and 26%. 24. An absorber rod for a pressurized water nuclear reactor, comprising:a tubular sheath and an absorber column formed by a stack of neutron absorbing elements received within the sheath, the absorber column containing at least two neutron absorbing elements, each neutron absorbing element being made of a material or a combination of materials selected from a group of neutron absorbing materials consisting of first europium hafnate, a second europium hafnate, a first samarium hafnate, a second samarium hafnate, hafnium carbide, and samarium hexaboride, the neutron absorbing elements differing in their cross-sectional structure before first use of the absorber rod in the pressurize water nuclear reactor, the first europium hafnate and the second europium hafnate being two defined compounds each formed by a reaction between hafnium oxide and europium oxide and having different compositions, the first samarium hafnate and the second samarium hafnate being two defined compounds each formed by a reaction between hafnium oxide and samarium oxide and having different compositions,the absorber rod being configured for being movable vertically during normal operation of the pressurized water nuclear reactor in order adjust a reactivity of the core of the pressurized water nuclear reactor. 25. An absorber cluster for a pressurized water nuclear reactor, comprising:a cluster of absorber rods, each absorber rod comprising a tubular sheath and an absorber column formed by a stack of neutron absorbing elements received within the tubular sheath, each of the neutron absorbing elements extending along a longitudinal axis and having a cross-sectional structure determined by a cross-sectional geometry, a composition and a porosity of the neutron absorbing element, the absorber columns containing at least two absorber assemblies and each of the absorber assemblies consists of a plurality of the neutron absorbing elements having a same cross-sectional structure,the cluster comprising at least a first absorber assembly of the absorber assemblies and a second absorber assembly of the absorber assemblies, the neutron absorbing elements respectively of the first absorber assembly and the second absorber assembly being formed from a same material or a same combination of materials selected from the group of neutron absorbing materials consisting of a first europium hafnate, a second europium hafnate, a first samarium hafnate, a second samarium hafnate, hafnium carbide and samarium hexaboride,the first europium hafnate and the second europium hafnate being two defined compounds each formed by a reaction between hafnium oxide and europium oxide, the first europium hafnate and the second europium hafnate having different compositions;the first samarium hafnate and the second samarium hafnate being two defined compounds each formed by a reaction between hafnium oxide and samarium oxide, the first samarium hafnate and the second samarium hafnate having different compositions,the cross-sectional structure of the neutron absorbing elements of the first absorber assembly being different from the cross-sectional structure of the neutron absorbing elements of the second absorber assembly before first use of the absorber cluster in the pressurized water nuclear reactorwherein the neutron absorbing elements of the first absorber assembly are made exclusively of neutron absorbing materials of the group consisting of the first samarium hafnate, the second samarium hafnate, the first europium hafnate and the second europium hafnate, while the neutron absorbing elements of the second absorber assembly are made exclusively of hafnium carbide. |
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abstract | A preclinical nuclear imaging detector system, including a gantry and one or more detector assemblies, each including a scintillator configured to interact with radiation emanating from a target test object being imaged and at least one pinhole collimator, having one or more pinhole apertures formed therein. The pinhole collimator is disposed between the target object and the scintillator, wherein a distance between the pinhole aperture and the scintillator is selected as a function of the number of pinhole apertures provided in the collimator and to optimize one of sensitivity or spatial resolution, such that the one or more pinhole apertures collectively project a unitary minified radiation image of the target object onto the scintillator. Further, one or more photosensors are optically coupled to the scintillator to receive interaction events from the scintillator. |
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summary | ||
abstract | A beam diffuser selector apparatus, system and method for use with a particle accelerator. A movable member has a plurality of beam diffusers mounted thereon, each having a different predetermined thickness. A driving device is coupled to the movable member. The driving device is configured to selectively move the movable member such that a selected one of the beam diffusers is positioned in a test position which is adjacent to an output of the particle accelerator and between the output of the particle accelerator and a device under test. A controller is coupled to the driving device. The controller has a user interface for receiving commands selecting a particular one of the plurality of beam diffusers and provides control signals to the driving device to selectively move the movable member such that the selected one of the plurality of beam diffusers is positioned in the test position. |
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claims | 1. A collimator ( 332 ) for a detector array ( 190 ) of an x-ray technique-based nonintrusive inspection apparatus ( 8 ), which includes: a body which includes: a support structure ( 332 ); and a plurality of septa ( 334 ) secured to the support structure, wherein center lines ( 346 ) of two of the septa located next to one another converge in a first direction ( 348 ), but surfaces ( 342 , 344 ) of the two septa facing one another diverge in the first direction. 2. The collimator ( 332 ) according to claim 1 , claim 1 wherein the support structure is made of a material having a first, lead component. 3. The collimator according to claim 2 , wherein the support structure includes first and second support members and the septa are secured between the first and second support members. claim 2 4. The collimator according to claim 1 wherein a lower tip of a first of the septa is spaced from a lower tip of a sixteenth of the septa by a first distance and an upper tip of the first septum is spaced from an upper tip of the sixteenth septum by a second distance that is approximately 1 mm less than the first distance. claim 1 5. The collimator according to claim 4 wherein the first distance is approximately 50 mm and the second distance is approximately 49 mm. claim 4 6. The collimator according to claim 1 wherein the material includes a first, lead component comprises at least 86 percent lead. claim 1 7. The collimator according to claim 6 wherein the material includes a second component which provides the material with a strength which is stronger than lead. claim 6 8. The collimator according to claim 7 wherein the second component includes tin. claim 7 9. The collimator according to claim 8 wherein the material comprises about 86 percent lead, 3 percent tin, and 11 percent antimony. claim 8 10. The collimator according to claim 1 wherein the body is molded. claim 1 11. An x-ray technique-based nonintrusive inspection apparatus ( 8 ) which includes: a tunnel section ( 14 ); an x-ray source ( 150 ) which, when operated, radiates at least a portion of the confines of the tunnel section; a detector array ( 190 ); and a collimator ( 332 ) mounted over the detector array, the collimator including a plurality of septa ( 334 ), with center lines ( 346 ) of two of the septa located next to one another converging in a direction ( 348 ) toward the x-ray source and surfaces ( 342 , 344 ) of the two septa that face one another do not converge towards the x-ray source. 12. The x-ray technique-based nonintrusive inspection apparatus according to claim 11 wherein the surfaces diverge from one another toward the x-ray source. claim 11 13. The x-ray technique-based nonintrusive inspection apparatus according to claim 11 wherein the septa is of a material that includes a first, lead component comprising at least 86 percent thereof. claim 11 14. The x-ray technique-based nonintrusive inspection apparatus according to claim 13 which includes a CT scanner subsystem ( 34 ) mounted for rotation relative to the tunnel section, the x-ray source forming part of the CT scanner subsystem and being rotatable therewith. claim 13 15. The x-ray technique-based nonintrusive inspection apparatus according to claim 14 wherein the material includes a second component which provides the material with a strength which is stronger than lead. claim 14 16. The x-ray technique-based nonintrusive inspection apparatus according to claim 15 wherein the second component includes tin. claim 15 17. The x-ray technique-based nonintrusive inspection apparatus according to claim 16 wherein the material comprises about 86 percent lead, 3 percent tin, and 11 percent antimony. claim 16 |
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claims | 1. A method for nondispersive neutron line focusing beyond the critical angle of mirrors comprising the step of:positioning an array of at least three supermirrors, at least two elliptical supermirrors and at least one flat supermirror, to deflect neutrons at a preferred angle so as to enable nondispersive neutron line focusing. 2. The method of claim 1, wherein said positioning step comprises:positioning said array of at least three supermirrors so that all supermirrors reflect at angles less than said critical angle so as to enable nondispersive neutron line focusing. 3. A method for nondispersive neutron point focusing beyond the critical angle of mirrors comprising the step of:positioning an array of at least two sets of three supermirrors, at least four elliptical supermirrors and at least two flat supermirrors, so that said first set of at least three supermirrors deflect neutrons vertically and said second set of at least three supermirrors deflect neutrons horizontally at a preferred angle so as to enable nondispersive neutron point focusing. 4. The method of claim 3, wherein said positioning step further comprises:positioning said array of at least two sets of three supermirrors, at least four elliptical supermirrors and at least two flat supermirrors, so that said first set of at least three supermirrors deflect neutrons vertically and said second set of at least three supermirrors deflect neutrons horizontally, and so that all supermirrors reflect at angles less than said critical angle so as to enable nondispersive neutron point focusing. 5. A method for nondispersive neutron line focusing beyond the critical angle of mirrors comprising the step of:positioning an array of at least three multilayer mirrors, at least two elliptical multilayer mirrors and at least one flat multilayer mirror, to deflect neutrons at a preferred angle so as to enable nondispersive neutron line focusing. 6. The method of claim 5, wherein said positioning step further comprises:positioning said array of at least three multilayer mirrors, at least two elliptical multilayer mirrors and at least one flat multilayer mirror, so that all multilayer mirrors reflect at angles less than said critical angle so as to enable nondispersive neutron line focusing. 7. A method for nondispersive neutron point focusing beyond the critical angle of mirrors comprising the step of:positioning an array of at least two sets of three multilayer mirrors, at least four elliptical multilayer mirrors and at least two flat multilayer mirrors, so that said first set of at least three multilayer mirrors deflect neutrons vertically and said second set of at least three multilayer mirrors deflect neutrons horizontally at a preferred angle so as to enable nondispersive neutron point focusing. 8. The method of claim 7, wherein said positioning step further comprises:positioning said array of at least two sets of three multilayer mirrors, at least four elliptical multilayer mirrors and at least two flat multilayer mirrors, so that said first set of at least three multilayer mirrors deflect neutrons vertically and said second set of at least three multilayer mirrors deflect neutrons horizontally, and so that all multilayer mirrors reflect at angles less than said critical angle so as to enable nondispersive neutron point focusing. 9. A method for nondispersive neutron line focusing beyond the critical angle of mirrors comprising the step of:positioning an array of at least three total external reflection mirrors, at least two elliptical total external reflection mirrors and at least one flat total external reflection mirror, to deflect neutrons at a preferred angle so as to enable nondispersive neutron line focusing. 10. The method of claim 9, wherein said positioning step further comprises:positioning said array of at least three total external reflection mirrors, at least two elliptical total external reflection mirrors and at least one flat total external reflection mirror, so that all total external reflection mirrors reflect at angles less than said critical angle so as to enable nondispersive neutron line focusing. 11. A method for nondispersive neutron point focusing beyond the critical angle of mirrors comprising the step of:positioning an array of at least two sets of three total external reflection mirrors, at least four elliptical total external reflection mirrors and at least two flat total external reflection mirrors, so that said first set of at least three total external reflection mirrors deflect neutrons vertically and said second set of at least three total external reflection mirrors deflect neutrons horizontally at a preferred angle so as to enable nondispersive neutron point focusing. 12. The method of claim 11, wherein said positioning step further comprises:positioning said array of at least two sets of three total external reflection mirrors, at least four elliptical total external reflection mirrors and at least two flat total external reflection mirrors, so that said first set of at least three total external reflection mirrors deflect neutrons vertically and said second set of at least three total external reflection mirrors deflect neutrons horizontally, and so that all total external reflection mirrors reflect at angles less than said critical angle so as to enable nondispersive neutron point focusing. |
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048620076 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT With reference now to FIG. 1, wherein like numerals designate like components throughout all the several figures, the thermal protection shell 20 of the invention is preferably used to protect a cask assembly 1 for carrying radioactive materials of different activities aboard a vehicle such as a tractor-trailer. In use, the thermal protection shell 20 covers the cask assembly 1 which may be mounted within a novel biaxial restraint cradle 3. The cradle 3 may in turn be secured onto the trailer of a tractor-trailer (not shown). Generally, the cask assembly itself has a cylindrical body 5 which is circumscribed on either end by toroidal impact limiters 7a, and 7b. Each of these impact limiters 7a, 7b is a donut-shaped shell of yieldable aluminum which is approximately one-fourth of an inch thick. Each of the toroidal impact limiters 7a, 7b is mounted around its respective end of the cylindrical body 5 by means of a support ring assembly 8a, 8b which in turn is secured to the cylindrical body 5 by a plurality of bolts 9. Disposed between the impact limiters 7a, 7b are a pair of opposing turnnions 11a, 11b and 11c, 11d. The two pairs of trunnions are disposed 180 degrees apart around the cylindrical body 5 of the cask assembly 1, and are receivable within two pairs of turn buckle assemblies 12a, 12b, and 12c, 12d (of which only 12a and 12b are visible) that form part of the cradle 3. The cylindrical body 5 is capped by a closure 13 at one end, and an end plate assembly 15 (shown in FIG. 3) at the other end. As is best seen in FIGS. 3 and 5, the cylindrical body 5 of the cask assembly 1 is generally formed by an outer container 18 which is surrounded by the thermal protection shell 20 on its exterior, and which contains in its interior one of two different shield inserts 22 or 23, depending upon the activity and type of radiation emitted by the material to be transported. While only two specific types of shield inserts 22 and 23 are specifically disclosed herein, it should be noted that the inserts 22 and 23 are merely exemplary, and that the cask assembly may in fact be used with any number of different types of shield inserts formed of different shielding materials and of different wall thicknesses for handling radioactive material within a broad range of activity and radiation type. When reference now to FIGS. 2A, 2B, and 2C, the thermal protection shell 20 which circumscribes the outer container 18 of the cask assembly 1 is formed from a pair of semi-cylindrical shell sections 24a, 24b which are rigidly interconnectable into thermal contact with one another. Each of the shell sections 24a, 24b includes a pair of cut-outs 26 (shown in FIG. 3) for admitting the trunnions 11a, 11b, 11c, and 11d. Each of the shell sections 24a, 24b is formed from a metal having a thermal coefficient of expansion which is greater than that of the metal that forms the walls of the outer container 18, and which is at least as heat-conductive as the metal which forms the walls 54 of the outer container 18. When the outer wall of the outer container 18 is formed from steel, the shell sections 24a, 24b are preferably formed from aluminum or magnesium or an alloy of either or both of these metals. The coefficient of thermal expansion of these metals is approximately twice that of the thermal coefficient of expansion of steel. Moreover, the high coefficient of thermal conductivity of each such metal insures that the thermal protection shell 20 will not significantly obstruct the conduction of decay heat conducted through the walls of the outer container 18 which is generated by the radioactive material held within the cask assembly 1. When the diameter of the outer container 18 is between forty and sixty inches, a wall thickness of approximately one-half of an inch is preferred for both of the shell sections 24a, 24b. Such a wall thickness renders the thermal protection shell 20, as a whole, thin enough to be conveniently retrofitted over many existing transportation casks without significantly adding to the weight thereof, yet is thick enough to maintain the structural integrity needed to expand away from the outer walls of the outer container when exposed to a source of intense thermal radiation, such as a fire. Finally, the preferred thickness of one-half of an inch provides enough mass to give the entire thermal protection shell 20 a significant latent heat of fusion, which will provide still more thermal protection through oblation should the cask 1 be exposed to intense heat. A plurality of top and bottom connecting assemblies 28, 29 are used to rigidly interconnect the two semi-cylindrical shell sections 24a, 24b. Since each of the connecting assemblies 28, 29 are identical in structure, a description will be made only of the top connecting assembly 28 circled in FIG. 2A. This connecting assembly 28 is formed from a pair of opposing semicircular lugs 30a and 30b which are integrally formed along the edges of the shell sections 24a and 24b respectively. These lugs 30a, 30b include mutually alignable bore holes 31a and 31b for receiving a connecting bolt 32. The threaded end 33 of the bolt 32 is engaged to a tension nut 34 as shown in FIG. 2B. The distance between the two lugs 30a, 30b (and hence the distance between the edges of the shell sections 24a, 24b) is largely determined by the extent of which the end 33 of the bolt 32 is threaded through the tension nut 34. A lock washer 35 is disposed between the tension nut 34 and the lug 30a to prevent the nut 34 from becoming inadvertently loosened. A pair of lock nuts 36a, 36b are threadedly engaged near the center portion of the connecting bolt 32 between the two lugs 30a and 30b. These lock nuts provide two functions. First, when properly adjusted, they prevent the tension nut 34 from applying excess tensile forces between the two shell sections 24a and 24b which might interfere with their expansion away from the outer container 18 in the event the cask assembly is exposed to a fire or other source of intense heat. Second, the nuts 36a, 36b eliminate all slack or play between the lugs 30a, 30b, thus insuring that the connecting assembly 28 rigidly interconnects the two shield sections 30a, 30b. Again, lock washers 37a, 37b are disposed between the lock nuts 36a and 36b and their respective lugs 30a and 30b to prevent any inadvertent loosening from occurring. An overlap 40 is provided between the edges of the two shell sections 24a and 24b to establish ample thermal contact and hence thermal conductivity between these shell sections. The overlap 40 is formed from an over flange 42 and recess 44 provided along the edge of shell section 24a which interfits with a complementary outer flange 46 and recess 48 provided along the opposing edge of shield section 24b. The actual length of the overlap 40 will vary depending upon the distance between the two lugs 30a and 30b as adjusted by the bolt 32, tension nut 34, and lock nuts 36a and 36b. In operation, the two sections 24a, 24b of the thermal protection shell 20 are installed over the cask assembly 1 by aligning the various cutouts 26a, 26b, 26c, and 26d with the corresponding trunnions of 11a, 11b, 11c, and 11d which project from the cylindrical body 5, and placing the sections 24a, 24b together so that the lugs 30a and 30b of each of the connecting assemblies 28, 29 are in alignment with one another and the flanges and recesses 42, 44, and 48, 46 of each overlaps 40 are interfitted. Next, the bolt 32, tension nut 35, lock nuts 36a, 36b, and lock washers 35, 37a, and 37b are installed in their proper positions with respect to the lugs 30a, 30b of each of the connecting assemblies 28, 29. The tension nut 34 is then screwed over the threaded end 33 of connecting bolt 32 until the interior surface of each of the shell sections 24a and 24b is pulled into intimate thermal contact with the outside wall 54 of the outer container 18. In the preferred method of installing the thermal protection shield, the tension nut 34 of each of the connecting assemblies 28, 29 is initially torqued to a selected maximum on the threaded shaft of the bolt 32 until the nut 34 imparts a significant tensile force between the two lugs 30a and 30b. This tensile force tends to squeeze the two shell sections 24a and 24b together around the outer wall 54 of the outer container 18 in a clamp-like fashion, which in turn removed any significant gaps between the outer surface of the wall 54 and the inner surface of the shell sections 24a and 24b by bending these sections into conformity with one another. In the next step, each of the nuts 34 is relaxed enough to prevent these tensile clamping forces from interfering with the expansion of the thermal protection shell 20 in the event of a fire, yet not so much as to cause the surfaces of the shell 20 and the outer container from becoming diengaged with one another. Thereafter, the lock nuts 36a and 36b are tightened against the faces of their respective lugs 30a and 30b to remove all slack in each connecting assembly 28, 29. The end result is a rigid interconnection between opposing edges of the shield sections 24a and 24b, wherein each of the opposing lugs 30a and 30b is tightly sandwiched between the tension nut 34 and lock nut 36a, or the head of the bolt 38 and lock nut 36b, respectively. If the outer container has no trunnions 11a, 11b, 11c, 11d, or other structural members which would prevent the surfaces of the shell 20 and outer container 18 from coming into intimate thermal contact, the shell 20 may assume the form of a tubular sleeve which may be, in effect, heat shrunk into contact over the container 18. This alternative method of installation comprises the steps of removing the impact limiters 7a, 7b, heating the shell to a temperature sufficient to radially expand it, sliding it over the wall 54 of the outer container 18, allowing it to cool and contract into intimate thermal contact with the wall 54, and reinstalling the impact limiters 7a, 7b. FIG. 2C illustrates the typical gap condition between the inner surface of the thermal protection shell 20 and the outer surface of the outer container 18. Under ambient conditions, these two opposing surfaces are either in direct contact with one another or separated by only a tiny gap 50 which may be as much as one mil. Such a one mil separation at various points around the cask assembly 1 does not significantly interfere with the conduction of heat between the wall 54 of outer cask 18, and the thermal protection shell 20. However, when the cask assembly 1 is exposed to a source of intense thermal radiation such as a fire, the substantially higher thermal coefficient of expansion of the aluminum or magnesium forming the shell 20 will cause it to expand radially away from the outer surface of the outer container 18, leaving an air gap 53 (shown in phantom) between the two surfaces. Moreover, since the thermal protection shield 20 is formed from a metal having good heat conductive properties, this differential thermal expansion is substantially uniform throughout the entire circumference of the shield 20, which means that the resulting insulatory air gap 53 is likewise substantially uniform. When this gap exceeds approximately two and one-half mils, the primary mode of heat transfer switches from conductive and convective to radioactive. Thus, the three mil gap provides a substantial thermal resistor between the fire and other source of intense infrared radiation in the outer container 18 of the cask 1. With reference now to FIGS. 3, 4A, 4B, and 5, the side walls of the outer container 18 of the improved cask 1 are a laminate formed from the previously mentioned outer wall 54, an inner wall 56, and a center layer 58 of shielding material. In the preferred embodiment, the outer wall 54 is formed from low alloy steel approximately one-fourth of an inch thick. Such steel is economical, easy to manufacture, and a reasonably good conductor of heat. In the alternative, stainless steel may be used in lieu of low alloy steel. While the use of stainless steel would be more expensive, it provides the additional advantage of corrosion-resistance. The inner wall 56 is preferably also formed from low alloy steel. However, the inner wall 56 is made two inches thick in order to provide ample structural rigidity and strength to the outer container 18. Disposed between the outer wall 54 and the inner wall 56 is a layer of Boro-Silicone. This material advantageously absorbs neutrons from neutron-emitting radioactive materials (such as transuranic elements), and further is a relatively good conductor of heat. It is a rubbery material easily cast, and may be melted and poured between the inner and outer walls 54, 56 of the outer container 18 during its manufacture. Boro-Silicone is available from Reactor Experiments, Inc., and is a registered trademark owned by that corporation. The bottom of the outer container 18 is formed by an end plate assembly 15 that includes an outer plate 60, an inner plate 62, and a layer of center shielding material 64. In the preferred embodiment, the outer plate 60 is again formed from a low alloy steel approximately one-forth inch thick. The inner plate 62, like the inner wall 56, is again formed from a layer of low alloy steel approximately two inches thick. The center shielding material 64 is again preferably Boro-Silicone for all the reasons mentioned in connection with the center shielding material 58 of the side walls of the container 18. The low alloy steel inner plate 62 is joined around the bottom edge of the inner wall 56a 360' via weld joint 66. The top of the outer container 18 includes a forged ring of low alloy steel 68. This ring 68 is preferably four inches thick throughout its length, and is integrally connected to the inner wall 56 of the container 18 by a 360' weld joint 69. The upper edge of the ring 68 is either threaded or stepped to accommodate one of the two types of improved closures 115b or 117b, as will be explained in detail hereinafter. With specific reference now to FIGS. 3 and 5, the cask assembly 1 is formed from the outer container 18 and shell 20 in combination with one of two different shield inserts 22 (illustrated in FIG. 3) or 23 (illustrated in FIG. 5). Each of the shield inserts 22, 23 is formed from an outer cylindrical wall 72 which is preferably one inch thick and a cylindrical inner wall 74 which is approximately one-fourth of an inch thick. Both walls are formed from A1S1 304 stainless steel. The corrosion resistance of stainless steel prevents the outer dimensions of the outer wall 74 from becoming distorted as a result of rust, which in turn helps advantageously to maintain a relatively tight, slack-free fit between the shield inserts 22, 23 and the interior of the outer container 18. Each of the shield inserts 22 and 23 includes a layer of shielding material 76 between their respective outer and inner walls 72, 74. However, in shield insert 22, this shielding material is formed from a plurality of ring-like sections 78a, 78b, and 78c of either depleted uranium or tungsten. These materials have excellent gamma shielding properties, and are particularly well adapted to contain and shield radioactive material emitting high intensity gamma radiation. Of course, a single tubular layer of depleted uranium or tungsten could be used in lieu of the three stacked ring-like sections 78a, 78b, and 78c. However, the use of stacked ring-like sections is preferred due to the difficulty of fabricating and machining these metals. To effectively avoid radiation streaming at the junctions between the three sections, overlapping tongue and groove joints 79 (see FIG. 4A) are provided at each junction. By contrast, in shield insert 23, a layer of poured lead 80 is used as the shielding material 76. While lead is not as effective a gamma shield as depleted uranium, it is a better material to use in connection with high-neutron emitting materials, such as the transuranic elements. Such high neutron emitters can induce secondary neutron emission when depleted uranium is used as a shielding material. While such a secondary neutron emission is not a problem with tungsten, this metal is far more difficult and expensive to fabricate than lead, and is only marginally better as a gamma-absorber. Therefore, lead is a preferred shielding material when high-neutron emitting materials are to be transported. It should be noted that the radius of the interior of the shield inserts 22 and 23 will be custom dimensioned with a particular type of waste to be transported so that the inner wall 74 of the insert comes as close as possible into contact with the radioactive material contained therein. The Applicant has noted that fulfillment of the foregoing criteria provides the most effective shielding configuration per weight of shielding material. Additionally, the thickness and type of shielding material 76 will be adjusted in accordance with the activity of the material contained within the shield insert 22, 23 so that the surface radiation of the cask assembly 1 never exceeds 200 mr. The fulfillment of these two criteria maximizes the capacity of the cask assembly 1 to carry radioactive materials while simultaneously minimizing the weight of the cask. FIGS. 4A and 4B illustrate the vent, purge, and drain assembly 90 of the outer container 18. This assembly 90 includes a threaded drain pipe 92 for receiving a drain plug 94. The inner end 96 of the drain plug 94 is conically shaped and seatable in sealing engagement with a complementary valve seat 97 located at the inner end of the pipe 92. Wrench flats 98 integrally formed at the outer end of the drain plug 94 allow the plug 94 to be easily grasped and rotated into or out of sealing engagement with the valve seat 97. A vent pipe 100 is obliquely disposed in fluid communication with the end of the drain pipe 92. A threaded vent plug 102 is engageable into and out of the vent pipe 100. A screw head 103 is provided at the outer end of the vent plug 102 to facilitate the removal or insertion of the threaded plug 102 into the threaded interior of the vent pipe 100. A drain tube 104 is fluidly connected at its upper end to the bottom of the valve seat 97 by way of a fitting 106. In the preferred embodiment, the drain tube 104 is formed from stainless steel, and is housed in a side groove 108 provided along the inner surface of the wall 56 of the outer container 18. As is most easily seen in FIG. 4B, the lower open end 109 of the drain tube 104 is disposed in a bottom groove 110 which extends through the shallowly conical floor 112 of the outer container 18. In operation, the vent, purge, and drain assembly may be used to vent the interior of the outer container 18 by removing the vent plug 102 from the vent pipe 100, screwing an appropriate fitting (not shown) into the threaded vent pipe 100 in order to channel gases to a mass spectometer, and simply screwing the conical end 96 of the drain plug 94 out of sealing engagement with the valve seat 97. If drainage is desired, both the drain plug 94 and vent plug 102 are again removed. Gas purging is preferably accomplished after draining by removing the vent plug 102, and connecting a source of inert gas to the drain pipe 92. The partial vacuum within the container 18 that created by the suction pump encourages inert gas to flow down through the drain tube 104. Although not specifically shown, the interior of the drain plug 98 may be provided with one or more rupture discs to provide for emergency pressure relief in the event that the cask assembly 1 is exposed to a source of intense thermal radiation, such as a fire, over a protracted period of time. A suction pump is connected to the drain pipe 92 in order to pull out, via drain tube 104, any liquids which may have collected in the bottom groove 110 of the conical floor 112 of the outer container 18. The closures 13 used in connection with the cask 1 may be either screw-type double-lidded closures 115a, 115b (illustrated in FIG. 3), or breech-lock double-lidded closures 117a, 117b (illustrated in FIG. 5). With references now to FIG. 3, each of the screw-type closures 115a, 115b includes an outer lid 120z, 120b, and an inner lid 112a, 112b. The inner lid 122a, 122b in turn includes an outer edge 124a, 124b which is seatable over a ledge 126a, 126b provided around the opening 128a, 128b of the shield insert 22 or the outer container 18 respectively. A gasket 130a, 130b circumscribes the outer edge 124a, 124b of each of the inner lids 122a, 122b of the two closures 115a, 115b. In the preferred embodiment, these gaskets 130a, 130b are formed of Viton because of its excellent sealing characteristics and relatively high temperature limit (392 degrees F.) compared to other elastomers. The gasket 130a, 130b of each of the inner lids 122a and 122b is preferably received and held within an annular recess (not shown) that circumscribes the outer edge 124a, 124b of each lid. Each of these gaskets 130a, 130b is capable of effecting a fluid-tight 360 degree seal between the outer edge 124a, 124b of each of the inner lids 122a, 122b and the ledges 126a, 126b. To facilitate the insertion of shield inset 22 into the container 18, it is important to note that the opening 128b of the container 18 is at least as wide as the interior of the container 18 at all points. Each of the outer lids 120a, 120b of the screw-type closures 115a, 115b includes a threaded outer edge 134a, 134b which is engageable within a threaded inner edge 136a, 136b that circumscribes the openings 128a, 128b of the shield insert 22 and the outer container 18 respectively. Swivel hooks 137a, 137b (indicated in phantom) may be detachably mounted to the centers of the outer lids 120a, 120b to facilitate the closure operation. Finally, both of the outer lids 120a, 120b of the screw-type closures 115a, 115b includes a plurality of sealing bolts 138a-h, 139a-h, threadedly engaged in bores extending all the way through the outer lids 120a, 120b for a purpose which will become apparent shortly. To seal the cask assembly 1, inner lid 122a is lowered over ledge 126a of the shield insert 22 so that the gasket 130 is disposed between the outer edge 124a of the inner lid 122a and ledge 126a. The detachably mountable swivel hook 137 is mounted onto the center of the outer lid 120a. The outer lid 120a is then hoisted over the threaded inner edge 136a of the shield insert 22. The threaded outer edge 134a of the outer lid 120a is then screwed into the threaded inner edge 136a to the maximum extent possible. The axial length of the screw threads 134a and 136a are dimensioned so that, after the outer lid 120a is screwed into the opening 128a to the maximum extent possible, a gap will exist between the inner surface of the outer lid 120a and the outer surface of the inner lid 122a. Once this has been accomplished, the securing bolts 138a-h are each screwed completely through their respective bores in the outer lid 120a so that they come into engagement with the inner lid 122a, thereby pressing the gasket 130a and into sealing engagement between the ledge 126a and the outer edge 124a of the lid 122a. The particulars of this last step will become more apparent with the description of the operation of the breech-lock double-lidded closures 117a, 117b described hereinafter. To complete the closure of the cask assembly 1, the outer screw-type closure 115b is mounted over the opening 128b of the outer container 18 in precisely the same fashion as described with respect to the opening 128a of the shield insert 22. With reference now to FIGS. 5, 6A, and 6B, the breech-lock double-lidded closure 117a, 117b also includes a pair of outer lids 140a, 140b which overlie a pair of inner lids 142a, 142b respectively. Each of the inner lids 142a, 142b likewise includes an outer edge 144a, 144b which seats over a ledge 146a, 146b that circumscribes the opening 148a, 148b of the shielding insert 23 and outer container 18, respectively. Each of the outer edges 144a, 144b is circumscribed by a gasket 150a, 150b for effecting a seal between the edges 144a, 144b and their respective ledges 146a, 146b. Like opening 128b, opening 148b is at least as wide as the interior of the outer container 18. Thus far, the structure of the breech-lock double-lidded closures 117a, 117b has been essentially identical with the previously described structure of the screw-type double-lidded closures 115a, 115b. However, in lieu of the previously described screw threads 134a, 134b, the outer edges 154a, 154b of each of the outer lids 140a, 140b are circumscribed by a plurality of uniformly spaced arcuate notches 156a, 156b which define a plurality of arcuate flanges 158a, 158b. Similarly, the inner edges 160a, 160b which circumscribe each of the openings 148a, 148b of the shield insert 23 and outer container 18, respectively, include notches 162a, 162b which also define arcuate flanges 164a, 164b. The flanges 158a, 158b which circumscribe each of the outer lids 140a, 140b are dimensioned so that they are insertable through the arcuate notches 162a, 162b which circumscribe the inner edges 160a, 160b of the shield insert 23 and the outer container 18. As may best been seen in FIG. 6A and 6C, such dimensioning allows the flanges 164a, 164b of each of the outer lids 140a, 140b, to be inserted through the notches 162a, 162b of each of the openings 148a, 148b and rotated a few degrees to a securely locked position wherein the arcuate flanges 158a, 158b of the outer lids 140a, 140b are overlapped and captured by the arcuate flanges 164a, 164b that circumscribe the inner edges 160a, 160b. It should be further noted that the axial length L1 (illustrated in FIG. 6B) of the interlocking flanges 158a, 158b and 164a, 164b is sufficiently short to leave a small gap L2 between the inner surface of the outer lids 140a, 140b and the outer surface of the inner lids 142a, 142b. The provision of such a small distance L2 between the outer and inner lids allows the outer lids 140a, 140b to be rotated a few degrees into interlocking relationship with their respective notched inner edges 160a, 160b without transmitting any rotary motion to the inner lids 142a, 142b which could cause the inner lid gaskets 150a, 150b to scrape or wipe across their respective ledges 146a, 146b. Connected around the outer edges of the outer lids 140a, 140b are three suspension pin assemblies 166a, 166b, and 166c and 167a, 167b, and 167c (not shown) respectively. Each of these suspension pin assemblies 166a, 166b, 166c and 167a, 167b, 167c are uniformly spaced 120' apart on the edges of their respective outer lids 140a, 140b. As the structure of each suspension pin assembly is the same, only a suspension pin assembly 166a will be described. With reference now to FIG. 6C, suspension pin assembly 166a includes a suspension pin 168 which is slideably movable along an annular groove 170 provided around the circumference of each of the inner lids 142a, 142b. A simple straight-leg bracket 172 connects the suspension pin 168 to the bottom edge of its respective outer lid. In operation, the suspension pin assemblies 166a, 166b, 166c and 167a, 167b, 167c, serve two functions. First, the three suspension pin assemblies attached around the edges of the two outer lids 140a and 140b mechanically connect and thus unitize the inner and outer lids of each of the breech-lock closures 117a, 117b so that both the inner and the outer lids of each of the closures 177a and 117b may be conveniently lifted and lowered over its respective opening 148a, 148b in a single convenient operation. Secondly, the pin-and-groove interconnection between the inner and the outer lids of each of the two breech-lock type closures 117a and 117b allows the outer lids 140a and 140b to be rotated the extent necessary to secure them to the notched outer edges 160a, 160b of their respective containers without imparting any significant amount of torque to their respective inner lids 142a, 142b. This advantageous mechanical action in turn prevents the gaskets 150a and 150b from being wiped or otherwise scraped across their respective ledges 146a, 146b. In the preferred embodiment, the width of the groove 170 is deliberately made to be substantially larger than the width of the pin 168 so that the pin 168 may avoid any contact with the groove 170 when the outer lids 140a, 140b are rotated into interlocking relationship with their respective containers 23 and 18. With reference again to FIG. 6A and 6C, each of the outer lids 140a, 140b includes eight sealing bolts 174a-h, 174.1a-h equidistantly disposed around its circumference. Each of these sealing bolts 174a-h, 174.1a-h is receivable within a bore 175 best seen in FIG. 6C. Each of these bores 175 includes a bottom-threaded portion 176 which is engageable with the threads 176.1 of its respective bolt 174a-h, 174.1a-h, as well as a centrally disposed, non-threaded housing portion 177. At its upper portion the bore 175 includes an annular retaining shoulder 178 which closely circumscribes the shank 179 of its respective bolt 174a-h, 174.1a-h. The retaining shoulder 178 insures that none of the sealing bolts 174a-h, 174.1a-h will inadvertently fall out of its respective bore 175 in the outer lid 140a, 140b. In operation, each of the sealing bolts 174a-h, 174.1a-h is screwed upwardly into its respective bore 175 until its distal end 179.1 is recessed within the threaded portion 176 of the bore 175. After the outer lid 140a or 140b has been secured into the notched inner edge 160a or 160b of its respective container 23 or 18, the sealing bolts 174a-h 174.1a-h are screwed down into the position illustrated in FIG. 6C until their distal ends 179.1 forcefully apply a downward-direction force around the outer edges 144 a, 144b of their respective inner lids 142a, 142b. Such a force presses the gaskets 150a and 150b into sealing engagement against their respective ledges 146a, 146b. It should be noted that the same bolt and bore configuration as heretofore described is utilized in the screw-type double-lidded closures 115a, 115b. To insure that the outer lids 140a and 140b will not become inadvertently rotated out of locking engagement with their respective vessels 23 or 18, a locking bracket 180 is provided in the position illustrated in FIG. 6A and 6B in each of the outer lids 140a, 140b after they are rotated shut. Each locking bracket 180 includes a lock leg 182 which is slid through mutually registering notches 156a, 156b, and 162a, 162b after the outer lids 140a and 140b have been rotated into locking engagement with the inner edges 160a, 160b of either the shielding insert 23 or the outer container 18. In the case of outer lid 140b, the mounting leg 184 is secured by means of locking nuts 186a, 186b. In the case of outer lid 140a, the mounting leg 184 is captured in place by inner lid 142b which abuts against it. Although not specifically shown in any of the drawings, each of the outer lids 120a, 120b of the screw-type double-lidded closures 115a, 115b is similarly secured. However, instead of a locking bracket 180, a locking screw (not shown) is screwed down through the outer edges of each of the outer lids 120a, 120b and into a recess precut in each of the inner lids 122a, 122b. |
description | This invention was made with Government support under Contract No. DE-NE0000583 awarded by the Department of Energy. The Government has certain rights in this invention. The following pertains to the nuclear reactor arts, nuclear power arts, nuclear reactor safety arts, and related arts. A typical nuclear reactor comprises a radioactive reactor core disposed in coolant in a lower portion of a reactor pressure vessel. For example, a light water reactor employs purified water as the coolant, and the reactor core typically comprises a uranium composition such as uranium oxide (UO2) enriched in the fissile 235U isotope. In operation, the nuclear reactor core supports a nuclear chain reaction in which radioactive decay events in the reactor core emit neutrons that stimulate additional decay events in the reactor core. The nuclear chain reaction generates heat that transfers to the coolant. In a boiling water reactor (BWR) design, heat from the reactor core converts coolant in the reactor pressure vessel to steam that is directly used to drive the turbine of an electrical generation system (or more generally to perform some other useful work). The BWR design has a disadvantage in that the steam piped from the reactor pressure vessel to the turbine has some contaminant radioactivity. In a pressurized water reactor (PWR) design, the coolant in the pressure vessel remains in a liquid state (e.g. subcooled state) and heats feedwater (secondary coolant) that flows through a separate flow path in a steam generator. The feedwater (secondary coolant) is converted to steam by heat transfer from the (primary) coolant of the reactor pressure vessel, thus providing steam for driving a turbine or performing other useful work that is free from contaminant radioactivity. In conventional PWR designs, the steam generator is separate from the nuclear reactor, and a (primary) coolant loop comprising large diameter piping flows primary coolant between the reactor pressure vessel and the separate steam generator (or generators). By contrast, in so-called “integral” PWR designs, the steam generator is located inside the reactor pressure vessel and feedwater is piped into the steam generator through suitable vessel penetrations. Advantageously, the integral PWR design avoids flowing primary coolant with its contaminant radioactivity through external large-diameter piping, and typical integral PWR designs reduce the diameter and number of vessel penetrations overall. In all such designs, the nuclear chain reaction in the reactor core generates high concentrations of neutrons in the reactor core. In a thermal nuclear reactor, the coolant also serves as a neutron moderator in order to thermalize neutrons to lower energies that are more effective for stimulating fissile isotope radioactive decay events. Neutron reflectors are typically disposed around the nuclear reactor core in order to retain a higher concentration of neutrons in the core. The neutron reflectors also serve to greatly reduce the neutron concentration outside the reactor pressure vessel. For example, in some designs it is expected that the neutron concentration external to the lower portion of the reactor pressure vessel (that is, proximate to the reactor core) is of order 100,000 lower than the concentration inside the reactor core. In some embodiments described herein as illustrative examples, a nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. In some additional embodiments the annular neutron stop comprises an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. In some such embodiments the inner neutron stop ring comprises a high temperature neutron absorbing material that is stable at an exterior temperature of the reactor pressure vessel during nuclear reactor operation, and the outer neutron stop ring comprises a neutron absorbing material that is different from the high temperature neutron absorbing material. In other embodiments the neutron absorbing materials are the same. In a more specific embodiment, the outer neutron stop ring includes: (i) an outermost neutron stop ring attached to the wall of the reactor cavity and comprising the neutron absorbing material that is different from the high temperature neutron absorbing material, and (ii) a middle neutron stop ring attached to the outermost neutron stop ring and comprising the high temperature neutron absorbing material. In some further embodiments described herein as illustrative examples, an apparatus as described in the immediately preceding paragraph further comprises a tube penetrating through the annular neutron stop, and a neutron plug disposed in the tube at the penetration of the tube through the neutron stop. The neutron plug comprises neutron absorbing material. The tube may comprise an excore instrument guide tube, and the apparatus may further comprise an excore instrument suspended from the neutron plug by a cable. In some still additional embodiments described herein as illustrative examples, a method comprises: disposing a lower portion of a nuclear reactor including a nuclear reactor core in a reactor cavity; generating a neutron field by operating the nuclear reactor; and retaining the neutron field in the reactor cavity. In some further embodiments the retaining comprises retaining the neutron field in the reactor cavity using an annular neutron stop located at an elevation above the nuclear reactor core. With reference to FIG. 1, a nuclear reactor includes a reactor pressure vessel 10 containing a nuclear reactor core 12. In diagrammatic FIG. 1, the location of the nuclear reactor core 12 inside the pressure vessel 10 is diagrammatically indicated, but it is to be understood that because the nuclear reactor core 12 is disposed inside the pressure vessel 10 it is not visible externally. The nuclear reactor core 12 is more particularly disposed in a lower portion of the reactor pressure vessel 10, which has the advantage that a loss of coolant accident (LOCA) is less likely to result in uncovering of the reactor core 12. Air exposure of the reactor core 12 is generally problematic as it can lead to deleterious air/core material reactions, melting of fuel rods, and so forth. To further reduce the likelihood of the reactor core 12 becoming uncovered, that is, exposed to air, the lower portion of the reactor pressure vessel 10 that contains the reactor core 12 is disposed in a reactor cavity 14 defined by a reactor cavity structure 15. In some embodiments, the reactor cavity structure 15 comprises a concrete (possibly steel-reinforced concrete) floor of a surrounding radiological containment structure (not shown). In the event of a LOCA, the reactor cavity 14 is suitably flooded with water, for example from a refueling water storage tank (RWST) 16 located inside the radiological containment via suitable piping 17. The RWST 16 is also used as a water source during refueling (that is, during replacement of some or all fuel assemblies that make up the reactor core 12). As further illustrative examples, the water source for flooding the reactor cavity 14 may be supplied from outside radiological containment, and/or may include (primary) reactor coolant that escapes from the reactor pressure vessel 10 into the radiological containment structure via the LOCA break. (Such coolant typically escapes the pressure vessel 10 in the form of steam or a steam/water mixture, which can be condensed and channeled into the reactor cavity 14 by suitable condensers, gutters, or the like located inside the radiological containment). The illustrative nuclear reactor is of the integral pressurized water reactor (integral PWR) variety, and further includes one or more internal steam generators 18 disposed inside an upper portion of the reactor pressure vessel 10, and reactor coolant pumps (RCPs) 20 mounted on the upper portion of the reactor pressure vessel 10. Like the nuclear reactor core 12, it is to be understood that the diagrammatically indicated steam generators 18 are disposed inside the pressure vessel 10 and hence are not actually visible externally. In the illustrative integral PWR, integral isolation valves (IIVs) 22 are employed at vessel penetrations, for example to inject feedwater into the steam generators 18 and to extract steam from the steam generators 18. The illustrative integral PWR is mounted in suspended fashion from at or near the top of the reactor cavity 14 via reactor support brackets, corbels or the like 24; alternatively, the integral PWR may be bottom-supported by a suitable support at the bottom of the reactor cavity 14. The illustrative integral PWR is an example, and the neutron stops and related subject matter disclosed herein are readily employed in conjunction with other types of nuclear reactors, such as PWR designs with external steam generators, with boiling water reactor (BWR) designs, and so forth. With continuing reference to FIG. 1, as previously mentioned neutron reflectors are typically disposed around the nuclear reactor core in order to retain a higher concentration of neutrons in the core, and the neutron reflectors also serve to greatly reduce the neutron concentration outside the reactor pressure vessel. Nonetheless, the neutron field generated by the reactor core 12 extends into the reactor cavity 14 outside of the reactor pressure vessel 10. During normal reactor operation, this region is not occupied by operations personnel, and does not contain radiation-sensitive equipment. Thus, the concentration of neutrons in the reactor cavity 14 outside of the reactor pressure vessel 10 may not be, in and of itself, problematic. However, as disclosed herein and as diagrammatically shown by arrows 26 in FIG. 1, a significant neutron field extends upward above the reactor cavity 14. Due to the reflectivity of neutrons, a substantial fraction of the neutrons that escape from the reactor vessel into the reactor cavity 14 scatter in the air, ricochet off of the inside walls of the reactor cavity 14 and travel generally upward to exit the top of the reactor cavity 14. In the illustrative integral PWR of FIG. 1, this exposes the reactor coolant pump (RCPs) 20 and the integral isolation valve (IIV) units 22 to high neutron densities, which can be problematic if these components are sensitive to neutron exposure, or even if these components are not rated for the neutron field level generated by this effect. While for the illustrative integral PWR exposure of the illustrative components 20, 22 is of concern, more generally a nuclear reactor tends to have potentially radiation-sensitive components such as electronics, components including polymer materials, or so forth located above the reactor cavity. Additionally, operations personnel may perform operations in this region, leading to undesirable human exposure to relatively high neutron radiation. With reference to FIG. 2, this effect of neutrons escaping upward out of the reactor cavity 14 is illustrated by a neutron field simulation performed for an integral PWR of the type shown in FIG. 1. In FIG. 2, region A corresponds to the nuclear reactor core 12 of FIG. 1, and as the source of the neutron field has the highest neutron density. A region O indicated in FIG. 2 corresponds to the reactor cavity structure 15 of FIG. 1, and is assumed to have a neutron concentration of zero (or a suitably negligible value for the modeling). Based on the neutron field modeling, it is found that the neutron field in a region B corresponding to the portion of the reactor cavity 14 just radially outside the pressure vessel 10 at the same elevation as the nuclear reactor core 12 has a neutron field about five orders of magnitude lower than the neutron concentration in the region A corresponding to the reactor core 12. This is still a relatively high neutron field. As further seen in FIG. 2, the neutron field has decreased by an additional factor of only 100 in a region C located directly above the open upper end of the reactor cavity 14, and in the remaining simulation volume D above the reactor cavity 14 the neutron field has decreased by only a factor of 1000 as compared with the neutron field in the region B. The high neutron field in the regions C, D is attributable to the neutron reflectivity of the steel making up the reactor pressure vessel 10 and of the concrete making up the reactor cavity structure 15, which combine to channel neutrons generally upward and into the regions C, D. In sum, it is recognized herein that the high neutron emissions of the reactor core 12 in combination with the reflective properties of neutrons and the presence of the reactor cavity 14 (which is typically required to ensure continued immersion of the reactor core 12 in the event of a LOCA) combine to channel neutrons generally upward into upper regions of radiological containment where sensitive equipment or operations personnel may be present. On the other hand, the reactor cavity structure 15 is typically a thick structure made principally of concrete, which prevents the passage of neutrons laterally or downward. Neutrons traveling inside the reactor pressure vessel 10 are typically thermalized by the coolant, so that this is also not a significant pathway for neutrons to escape upward. With returning reference to FIG. 1, it is disclosed herein that in view of the foregoing a sufficient neutron barrier is provided by an annular neutron stop 30 located in an upper region of the reactor cavity 14. The annular neutron stop 30 comprises a neutron-absorbing material filling an annular gap between the reactor pressure vessel 10 and the reactor cavity 14, and is effective to block neutrons from passing upward through the neutron stop 30. (It should be understood that the neutrons diagrammatically indicated by arrows 26 in FIG. 1 are provided for expository purpose, and are shown assuming that the neutron stop 30 is not present). The annular neutron stop 30 is located at an elevation that is above the uppermost elevation of the nuclear reactor core 12, that is, above the top of the nuclear reactor core 12. The annular neutron stop 30 is a ring of neutron shielding material that fills the annular gap between the reactor vessel 10 and the (inside) wall of the reactor cavity 14. The annular neutron stop 30 can be a continuous ring of shielding material, or made from segmented portions for installation purposes. With reference to FIG. 3, an example of a segmented embodiment is shown. The neutron stop 30 of FIG. 3 is assembled along the circumferential direction of blocks 32 of neutron-absorbing material. The joinder of each pair of neighboring blocks 32 preferably includes a stair-stepped or staggered interface 34 which eliminates direct line-of-site passageways so as to block neutrons from streaming through the interfaces 34 between the blocks 32. While the illustrative approach of employing blocks 32 along the circumferential direction is advantageous for assembly convenience, it is also contemplated for the neutron stop to be manufactured as a continuous ring rather than being segmented into blocks 32. With reference to FIG. 4, radial thermal expansion of the reactor pressure vessel 10 and radial expansion from interior pressure is expected to occur as the nuclear reactor is transitioned from standby into operation. Additionally, a large thermal gradient is present as the portion of the neutron stop 30 proximate to the reactor pressure vessel 10 is exposed to higher temperature as compared with the outboard portion of the neutron stop 30 that is proximate to the wall of the reactor cavity 14. One approach for accommodating these factors is illustrated in FIG. 4, where the neutron stop 30 is constructed as three component rings: an outermost neutron stop ring 301 that is secured to the wall of the reactor cavity 14; a middle neutron stop ring 302 located inboard of and secured to the neutron stop ring 301; and an innermost neutron stop ring 303 that is secured to the exterior of the reactor pressure vessel 10. Each component neutron stop ring 301, 302, 303 may be constructed along the circumferential direction using the illustrative blocks 32 with stair-stepped or staggered interfaces 34 between the blocks, as described with reference to FIG. 3. This same staggering may optionally exist as between the radial interface of the component rings. Additionally, an interface 36 between the middle and innermost neutron stop rings 302, 303 may be stair-stepped or staggered as shown in FIG. 4, to block neutrons from streaming through the interface 36 between the middle and innermost neutron stop rings 302, 303. In illustrative FIG. 4 this is obtained by including a cutout portion in the middle neutron stop ring 302 into and over which is disposed the innermost neutron stop ring 303. The interface between the outermost and middle neutron stop rings 301, 302 is not staggered in the illustrative example of FIG. 4, because the middle neutron stop ring 302 is bonded directly to the outermost neutron stop ring 301 with no air gap; however, it is also contemplated to employ a stair-stepped or staggered interface between the outermost and middle neutron stop rings. The blocks 32 that form the innermost neutron stop ring 303 are suitably attached to the exterior of the reactor pressure vessel 10 by welding, fasteners (e.g. bolts) or are held onto the vessel 10 with band straps. The innermost neutron stop ring 303 is allowed to grow radially with the reactor. To accommodate this, the interface 36 should have a sufficient air gap to accommodate differential thermal expansion, that is, to accommodate the outward expansion of the innermost stop ring 303 as the reactor pressure vessel 10 is heated to its operational temperature. Because the innermost neutron stop ring 303 is in contact with the reactor pressure vessel 10, it is exposed to heat emanating from the operating nuclear reactor via radiative, conductive and conductive thermal transfer pathways. This can be accommodated by containing the neutron shielding material of the innermost neutron stop ring 303 within an outer can comprising stainless steel or another suitably heat resistant structural material. In this case, the neutron-absorbing material does not need to be self-supporting or be of structural material. The outboard portion comprising the middle neutron stop ring 302 and outermost neutron stop ring 301 is exposed to lower temperature, but there is a substantial temperature gradient. This is accommodated by constructing the outboard portion of the neutron stop using two rings: the middle neutron stop ring 302 and the outermost stop ring 301. This construction allows the middle neutron stop ring 302 to be made of relatively more heat resistant material as compared with the outermost neutron stop ring 301. In some embodiments, both the innermost neutron stop ring 303 and the middle neutron stop ring 302 are made of the same material, while the outermost neutron stop ring 301 is made of a less heat resistant material that may be of lower cost and/or improved structural strength. The neutron stop rings 301, 302, 303 are suitably made of neutron-absorbing material that also has appropriate thermal insulation characteristics so that the neutron stop ring 30 does not act as a thermal shunt. It is also recognized herein that it is advantageous for the neutron stop material to include a neutron moderator material so as to thermalize neutrons which increases effectiveness of the neutron absorber. Some suitable materials are composite materials including a boron-containing neutron absorber component, a component including carbon and/or hydrogen which acts as a neutron moderator, and a thermally insulating matrix material. For the high temperature material of the inner neutron stop rings 302, 303, some suitable compositions include boron carbide (B4C) powder as the neutron absorber, disposed in a vermiculite matrix which provides both neutron moderator and thermal insulator characteristics. Vermiculite is a hydrous silicate mineral material, and the hydrogen provides neutron moderation. Serpentine is another hydrous thermally insulating material that may be substituted for vermiculite. As already noted, this material may be contained within an outer can comprising stainless steel or another suitably heat resistant structural material. The outermost neutron stop ring 301 is suitably made of borated concrete of suitably high density, optionally infused with hydrogen to enhance its neutron moderator characteristics. Since the outermost neutron stop ring 301 is structural and indeed supports the middle neutron stop ring 302 in a cantilevered fashion, it is contemplated for the borated concrete to be steel-reinforced or otherwise structurally reinforced. In another suitable embodiment, the outermost neutron stop ring 301 is made of borated polyethylene, which is advantageously lighter in weight as compared with concrete. The annular neutron stop 30 is located at an elevation that is above the uppermost elevation of the nuclear reactor core 12, and comprises neutron absorbing material that fills the annular gap between the reactor pressure vessel 10 and the reactor cavity 14. The annular neutron stop 30 is thus positioned to block neutrons from passing upward through the annular gap between the reactor pressure vessel 10 and the reactor cavity 14 and out into the space above the reactor cavity 14 which may be occupied by neutron radiation-sensitive equipment such as the RCPs 20 or IIV's 22, and/or which may be occupied by operational personnel. With reference to FIGS. 5 and 6, the reduction in neutrons escaping upward out of the reactor cavity 14 achieved by use of the disclosed neutron stop 30 is illustrated by neutron field simulations similar to that of FIG. 2, but now including simulated embodiments of the neutron stop 30 in a region S indicated in FIGS. 5 and 6. In FIG. 5 the simulated neutron stop includes: (i) an inner ring of a high temperature neutron-absorbing composite comprising 50-vol % B4C, 25-vol % vermiculite, and 25 vol-% air (modeling the impact of air pockets or the like in the composite material); and (ii) an outer ring of borated polyethylene. In FIG. 6, the simulated neutron stop includes: (i) an inner ring of 50-vol % B4C and 50-vol % air (modeling an inner ring comprising boron carbide powder in a steel encasement); and (ii) an outer ring of borated concrete. It is seen in FIGS. 5 and 6 that the neutron field in the reactor core 12 (that is, in region A) and in the region B corresponding to the portion of the reactor cavity 14 just radially outside the pressure vessel 10 at the same elevation as the nuclear reactor core 12 are not significantly affected by the presence of the simulated neutron stop 30. Note that in FIGS. 5 and 6 the region B is located below the region S of the simulated neutron stop. By contrast, the neutron field in the regions C, D located above the top of the reactor cavity 14 (and hence above the region S of the simulated neutron stop in the simulations of FIGS. 5 and 6) is greatly reduced as compared with the simulation of FIG. 2 which does not simulate the neutron stop 30. In the region C directly above the reactor cavity 14, the neutron field is reduced by a factor of 106 (that is, by a factor of one million) in the simulations of FIGS. 5 and 6 as compared with the simulation of FIG. 2. In the simulation volume D the neutron field has again decreased by a factor of 106 (that is, by a factor of one million) in the simulations of FIGS. 5 and 6 as compared with the simulation of FIG. 2. It should be noted that the simulated neutron stops of FIGS. 5 and 6 do not account for the detailed construction of the illustrative neutron stop 30, but rather model the two high-temperature neutron stop rings 302, 303 as a single ring of high temperature material. In particular, the simulations of FIGS. 5 and 6 do not model the air gap 36. However, the air gap 36 is at its smallest extent during normal operation of the integral PWR since during normal operation the reactor pressure vessel 10 is at temperature and hence maximally thermally expanded outward. As a further mechanism for reducing the neutron field in the reactor cavity 14, it is contemplated to line the reactor cavity walls with a surface layer of borated (and optionally hydrogen-infused) concrete. (This optional aspect is not incorporated into the simulations of FIGS. 2, 5, and 6). The neutron stop ring 30 should be designed so as to not impede operation of the reactor cavity 14. For example, the neutron stop 30 should not impede flooding of the reactor cavity 14 during a LOCA. Advantageously, piping 17 extending between the reactor cavity 14 and the RWST 16 is not expected to be a substantial neutron leak source since neutrons channeled into the RWST 16 by the piping 17 are expected to be removed by water in the RWST 16; nonetheless, if desired a neutron-absorbing check valve 38 (see FIG. 1) may be provided within the portion of the pipe 17 passing through the concrete wall of the reactor cavity 14 to block neutron leakage via pipe 17. In a suitable embodiment, the check valve 38 can comprise a conventional swing check valve allowing flow only in the direction into the reactor cavity 14, in which the valve disk comprises a neutron absorbing material such as a material comprising a boron compound and an optional neutron moderator material. The discharge end of the piping 17 should output into the reactor cavity 14 at an elevation that is below the elevation of the neutron stop 30. Condensed steam from the LOCA break can be channeled into the reactor cavity 14 via the staggered gap 36 between the neutron stop rings 302, 303. Another consideration is the potential for water flooding into the reactor cavity 14 to evaporate and form steam, potentially leading to a rise in pressure. This also can be accommodated by the staggered air gap 36 between the neutron stop rings 302, 303, which serves as a steam vent. With reference to FIG. 7, excore instrument guide tubes, such as the illustrative excore guide tube 40, present another potential neutron leak. Excores are instrumentation that monitors aspects such as the power distribution of the reactor. The excore guide tubes 40 are located beyond the radial profile of the reactor vessel, and as seen in FIG. 7 may be arranged to pass through the neutron stop ring 30 (and more particularly the outermost neutron stop ring 301 in FIG. 7). The guide tube 40 allows for the excore detector instrument to be lowered into the reactor cavity to an elevation equal to the height of the core. Due to the requirement to monitor power levels of quadrants of the core, the excores are clocked at specific locations within the reactor cavity, and accordingly their location may be difficult or impossible to alter in a given nuclear reactor design. With continuing reference to FIG. 7 and with further reference to FIGS. 8 and 9, a neutron plug 42 is provided at the penetration of the guide tube 40 through the neutron stop 30 in order to block neutrons from leaking through that penetration. An excore instrument 44 is suspended from the neutron plug 42 by a cable 46 whose upper end is connected to the bottom of the neutron plug 42. The neutron plug 42 is made of a suitable neutron-absorbing material, and in some embodiments may be made of the same material as the neutron stop through which the guide tube 40 penetrates. In the illustrative example, the guide tube 40 includes a protrusion or ledge 47 that engages a corresponding step 48 on the neutron plug 42 so that the neutron plug 42 rests on the protrusion or ledge 47 to keep it at the desired elevation corresponding to the penetration through the neutron stop 30. Additionally, the larger diameter of the neutron plug 42 above the step 48 blocks neutrons that might otherwise pass through the gap between the smaller diameter of the neutron plug 42 and the larger diameter of the excore guide tube 40. An electrical wire or wire bundle (or electrical cable, et cetera) 50 carries signals and/or power to/from the excore instrument 44. As seen in the cutaway view of the neutron plug 42 shown in FIG. 9, to allow the wire(s) 50 to pass through the neutron plug 42 without presenting a neutron leakage path, the wires(s) 50 pass through a torturous path through the neutron plug 42 that does not allow neutrons to bypass the shielding material of the neutron plug 42 and pass through the electrical cable 50. The passage of the wire(s) through the tortuous path of the neutron plug 42 can be achieved by forming a tortuous conduit and then passing the wire(s) 50 though the conduit. This allows the wire(s) 50 to be a continuous length passing through the neutron plug 42 and extending above and below it. In an alternative approach, the neutron plug 42 is formed by overmolding over a length of wire(s), and the wires above and below the neutron plug are then soldered or otherwise connected to the ends of the length of wire(s) molded into the neutron plug. With continuing reference to FIGS. 7-9 and with further reference to FIGS. 10 and 11, in the illustrative embodiment a lifting eye 52 is located at the top of the neutron plug 42. As best seen in FIG. 10, the lifting eye 52 can be connected to a lifting cable 54 so that the assembly can be raised and lowered inside the guide tube 40. FIG. 11 shows the lifting configuration in the context of the guide tube 40. With reference back to FIG. 7, in some embodiments the excore guide tube 40 may include a portion 56 extending downward and into a basement or other location at which instrument electronics, control, or the like are located. Optionally, an end 58 of this extension is similarly plugged, such as using another instance of the neutron plug 42 (not shown), so as to block neutron leakage into that location via the excore guide tube 40. While illustrated in the context of a penetration for the illustrative excore guide tube 40 through the neutron stop 30, the illustrative neutron plug 42 is readily modified to block neutron leakage at other types of penetrations through the neutron stop 30, such as for power cables in the case of alternative nuclear reactor designs having reactor coolant pumps located in the reactor cavity. Illustrative embodiments including the preferred embodiments have been described. While specific embodiments have been shown and described in detail to illustrate the application and principles of the invention and methods, it will be understood that it is not intended that the present invention be limited thereto and that the invention may be embodied otherwise without departing from such principles. In some embodiments of the invention, certain features of the invention may sometimes be used to advantage without a corresponding use of the other features. Accordingly, all such changes and embodiments properly fall within the scope of the following claims. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the present disclosure be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. |
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claims | 1. A method of passively cooling a spent fuel pool in a nuclear power plant absent of onsite and offsite power, wherein the spent fuel pool has an inner wall and an outer wall, the method comprising:spacing the inner wall and the outer wall of the spent fuel pool a distance apart substantially along a periphery of the spent fuel pool forming a gap,wherein, the gap contains air to impede a flow of heat from the spent fuel pool when passive cooling is deactivated;providing a heat sink;providing one or more conductive members having a first end and a second end;connecting the first end of the one or more conductive members to the gap and the second end to the heat sink; andactivating passive cooling of the spent fuel pool, comprising:adding water to substantially fill the gap;conducting heat generated from the spent fuel pool through the gap; andtransporting the heat from the gap through the one or more conductive members to the heat sink. 2. The method of claim 1, wherein adding water to substantially fill the gap, comprises:obtaining a water source;connecting a discharge header to the water source;discharging water from the water source into the discharge header; anddischarging the water from the discharge header into the gap. 3. The method of claim 2, further comprising installing in the discharge header a flow control valve having an open position and a closed position. 4. The method of claim 2, further comprising locating the discharge header at the top or near the top of the gap. 5. The method of claim 1, further comprising partitioning the gap into a plurality of channels. 6. The method of claim 5, further comprising installing in each of said plurality of channels a discharge header. |
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046845046 | claims | 1. In a fuel assembly for use at non-control rod locations of a nuclear reactor core, said fuel assembly including top and bottom nozzles and a plurality of longitudinal structural members extending between and attached to said nozzles for forming said assembly into an integral unitary structure, at least certain of said structural members including an elongated hollow cladding tube extending between said top and bottom nozzles and means secured to opposite ends of said tube for hermetically sealing said tube and attaching it to said top and bottom nozzles, the improvement which comprises: (a) a quantity of irradiation-induced creep resistant material disposed within said tube; and (b) pretensioning means positioned within said tube for applying a predetermined axially-directed compressive load to said creep resistant material therein and reacting said load so as to axially preload said tube in a state of pretension having a magnitude sufficient to substantially counteract an axial load typically transmitted through said unitary structure of said fuel assembly and thereby greatly reduce the compressive stress in said tube of said structural member. (a) an elongated hollow cladding tube extending between said top and bottom nozzles; (b) means secured to opposite ends of said tube for hermetically sealing said tube and attaching it to said top and bottom nozzles; (c) a quantity of irradiation-induced creep resistant material disposed within said tube, said creep-resistant material being a ceramic material in pellet form and coated with a burnable absorber material; and (d) pretensioning means positioned within said tube for applying a predetermined axially-directed compressive load to said creep resistant ceramic pellet stack therein and reacting said load so as to axially preload said tube in a state of pretension having a magnitude sufficient to substantially counteract an axial load typically transmitted through said unitary structure of said fuel assembly and thereby greatly reduce the compressive stress in said tube of said structural member. 2. The longitudinal structural member as recited in claim 1, wherein said creep-resistant material is a ceramic material in pellet form. 3. The longitudinal structural member as recited in claim 2, wherein said ceramic material is zirc oxide. 4. The longitudinal structural member as recited in claim 2, wherein said ceramic material is coated with a burnable absorber material. 5. The longitudinal structural member as recited in claim 1, wherein said pretensioning means is an elongated bellows type device positioned within said tube between said creep resistant material and one of said tube ends, said interior of said bellows type device being pressurized to create a predetermined axial force therein which places said creep resistant material in compression and said tube in said state of pretension. 6. The longitudinal structural member as recited in claim 5, wherein said pretensioning means further includes pressurization of the remainder of said tube. 7. The longitudinal structural member as recited in claim 1, wherein said pretensioning means is an arrangement of belleville springs positioned within said tube between said creep resistant material and one of said tube ends so as to create a predetermined axial force therein which places said creep resistant material in compression and said tube in said state of pretension. 8. The longitudinal structural member as recited in claim 7, wherein said belleville springs in said arrangement thereof are both stacked in parallel and in series. 9. In a fuel assembly for use at non-control rod locations of a nuclear reactor core, said fuel assembly including top and bottom nozzles and a plurality of longitudinal structural members extending between and attached to said nozzles for forming said assembly into an integral unitary structure, at least certain of said structural members comprising: 10. The longitudinal structural member as recited in claim 9, wherein said pretensioning means is an elongated bellows type device positioned within said tube between said creep resistant ceramic pellet stack and one of said tube ends, said interior of said bellows type device being pressurized to create a predetermined axial force therein which places said creep resistant ceramic pellet stack in compression and said tube in said state of pretension. 11. The longitudinal structural member as recited in claim 10, wherein said pretensioning means further includes pressurization of the remainder of said tube. 12. The longitudinal structural member as recited in claim 9, wherein said pretensioning means is an arrangement of belleville springs positioned within said tube between said creep resistant ceramic pellet stack and one of said tube ends so as to create a predetermined axial force therein which places said creep resistant ceramic pellet stack in compression and said tube in said state of pretension. 13. The longitudinal structural member as recited in claim 12, wherein said belleville springs in said arrangement thereof are both stacked in parallel and in series. |
claims | 1. A method to load a nuclear fuel rod, comprising:providing an apparatus for loading the nuclear fuel rod, the apparatus comprising:a fuel plate transfer unit, having an underside, the underside having openings for nuclear fuel pellets;an interfacing surface on which nuclear fuel pellets in the openings of the underside of the fuel plate transfer unit can roll; anda rotational belt positioned to accept nuclear fuel pellets that fall from the interfacing surface and the underside of the fuel plate transfer unit, the rotational belt positioned to transfer the nuclear fuel pellets into a fuel rod cladding open on both ends;providing nuclear fuel pellets housed in the openings in the underside of the fuel plate transfer unit, wherein the openings are sized to allow the nuclear fuel pellets to roll along the interfacing surface while keeping the nuclear fuel pellets from interacting with one another;transferring the nuclear fuel pellets from the fuel plate transfer unit in a fuel pellet column onto the rotational belt;indexing the nuclear fuel pellets in the fuel pellet column to a nuclear fuel pellet loading machine; andtransferring the fuel pellet column from the rotational belt into the fuel rod cladding. 2. The method to load a nuclear fuel rod according to claim 1, wherein the transferring of the nuclear fuel pellets from the fuel plate transfer unit to the rotational belt is through a slot in the interfacing surface. 3. The method to load a nuclear fuel rod according to claim 1, wherein the openings and the interfacing surface are configured as smooth surfaces. 4. The method to load a nuclear fuel rod according to claim 1, wherein the transferring of the nuclear fuel pellets from the fuel plate transfer unit to the rotational belt is by indexing a row of the fuel pellets in the fuel plate transfer unit over a slot in the interfacing surface over the rotational belt, and transferring the column of fuel pellets onto the rotational belt through the slot in the interfacing surface. 5. The method to load a nuclear fuel rod according to claim 1, wherein the transferring of the nuclear fuel pellets from the rotational belt into the fuel rod cladding is through a fuel element transfer apparatus. 6. The method to load a nuclear fuel rod according to claim 1, further comprising:verifying a length of the fuel pellet column after the step of indexing the nuclear fuel pellets in the fuel pellet column to a nuclear fuel pellet loading machine. 7. The method to load a nuclear fuel rod according to claim 1, further comprising: welding a lower end plug on the lower end of the cladding. 8. The method to load a nuclear fuel rod according to claim 7, further comprising: welding a top end cap on the top end of the cladding. |
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abstract | This invention generally concerns radioactive decontamination of deposits on components in a nuclear power plant and is specifically concerned with improved compositions, systems and methods for disrupting, dissolving, removing and reducing at ambient temperature radionuclides formed on the primary side surfaces of components in a pressurized water reactor and the internal components of a boiling water reactor. The methods include identifying the structure, taking the structure out of operational service, contacting the structure with an aqueous solution (e.g., a recirculating flow or static immersion), and adding an effective amount of elemental metal in solid form to the aqueous solution. |
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