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claims | 1. An accumulator provided with a flow damper inside, the flow damper including a cylindrical vortex chamber, a small flow pipe connected to a peripheral portion of the vortex chamber along a tangential direction thereto, a large flow pipe connected to the peripheral portion while forming a predetermined angle with this small flow pipe, and an outlet pipe connected to an outlet formed at a central part of the vortex chamber, whereinthe flow damper comprises a colliding jet controller for controlling a colliding jet composed of a jet from the large flow pipe and a jet from the small flow pipe flowing into the vortex chamber at the time of a large flow injection so that the colliding jet may proceed directly to the outlet without forming a vortex in the vortex chamber, the colliding jet controller is a bevel formed at a junction of the small flow pipe and the vortex chamber, andsaid bevel is formed in a predetermined shape having a predetermined surface, in which said predetermined surface extends from an inner surface, which is located at the opposite side of the large flow pipe, of the small flow pipe to an inner peripheral surface of the vortex chamber. 2. The accumulator according to claim 1, whereinthe bevel formed at the junction being in an oblique direction relative to the direction of the jet from the small flow pipe. 3. The accumulator according to claim 1, whereinthe bevel formed at the junction being bent. |
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044915403 | description | Referring first to FIG. 1, a number of spent nuclear fuel rods 11 from a nuclear reactor are arranged in a copper container 10. The fuel rods, which consist of zircaloy cladding tubes containing pellets of uranium dioxide, remaining attached to spacers 12 which retained the fuel rods in bundles in the nuclear reactor. These spacers 12 can be of stainless steel. In FIGS. 1 and 2, four fuel rod bundles 13, 14, 15 and 16 are shown. The fuel rod bundles may possibly rest on supports (not shown) spacing them from the bottom of the container 10 or they can be placed on a bed of copper powder. The container 10 is then filled in its entirety, while being vibrated, with a mixture 17 consisting of 70 parts by weight of a copper powder with spherical particles having diameters in the range 0.5-1.5 mm and of 30 parts by weight of a copper powder with spherical particles having diameters in the range 0.1-0.2 mm. A lid 18 of copper is then placed on the container 10. The container, the lid and the powder are all of the previously mentioned copper quality containing 99.95% Cu (including small amounts of Ag). The circumferential part 19 of the lid 18, which makes contact with the container 10, has a stepped shape to provide a central lower portion 20 of the lid which projects into the container. The confronting surfaces 10a and 18a of the container 10 and the lid 18, respectively, are roughened or otherwise textured, as is indicated in FIG. 3. The surfaces 10a and 18a are well cleaned and freed from oxide by acids before fitting the lid 18 onto the container 10. The container 10, its contents 11, 12, 17 and the lid 18 are arranged in a capsule 21 of copper sheet or of steel sheet, the lid 22 of which, made of copper sheet or steel sheet is welded to the capsule by forming a gas-tight joint 23. The lid 22 is provided with a tube 24 of copper or steel, respectively, which can be connected to a vacuum pump for evacuation of the capsule with its contents. After evacuation, the capsule is sealed by closing the tube 24 above the upper surface of the lid (e.g. by cold or hot welding). The sealed capsule 21, 22 with its contents is then subjected to hot isostatic pressing in two stages employing a gas, for example argon, as the pressure medium in a high pressure furnace of the kind disclosed in U.S. Pat. No. 4,172,807. In the first stage, the capsule is subjected to a pressure of 80 MPa and to a temperature of 450.degree.-500.degree. C. for a period of 2-10 hours. During the first stage, the copper in the container 10, the lid 18 and the powder 17 undergo a creep deformation, which results in the copper filling powder 17 providing an efficient all-round support for the fuel rods 11, which prevents creep rupture in the zircaloy cladding tubes as a result of an increase in pressure of the gas, present in these tubes, during continued heating. However, this first stage does not result in the powder grains, the container and the lid forming a coherent unit with a fully developed bonding. Such a result is achieved during the second stage in which the temperature in the furnace is increased to about 700.degree. C., while the pressure is increased, without additional supply of gas, to about 100 MPa, and by maintaining these conditions for 1-4 hours. When the capsule with its contents has been subjected to the second stage of the isostatic pressing, the capsule with its contained material is allowed to cool, whereafter the pressure is reduced to atmospheric pressure and the capsule is removed from the furnace. Normally, the capsule is allowed to remain around the compressed product 10, 11, 12, 17, 18 when it is to be deposited for long-term storage. In an alternative example, the mixture 17 consists of 55 parts by weight of a copper powder with spherical particles having diameters in the range 0.8-1.0 mm and 45 parts by weight of a copper powder with spherical particles having diameters in the range 0.2 mm and below. A fill density of 81% of the theoretical density can then be obtained by vibrational filling. After evacuation of the capsule 21 with its contents, the capsule is heated to 350.degree. C., whereupon it is filled with hydrogen gas with a pressure of 0.1 MPa. When this temperature has been maintained for 1/2 hour, the capsule is re-evacuated and is then refilled with hydrogen gas. This treatment with hydrogen gas at 350.degree. C. is repeated a plurality of times, for example 7 times, suitably with a successively longer treatment time after each refilling. The final treatment time could be 10 hours. The cyclic treatments with hydrogen gas result in a reduction of possibly existing oxides of copper. After completion of the cyclic treatments with hydrogen gas, the capsule 21, 22 is evacuated and sealed as in the previously described case. During the isostatic pressing, a temperature of 400.degree.-450.degree. C. is used in the first stage and a temperature of 525.degree. C. is used in the second stage. This described alternative example is otherwise carried out under the same conditions as the previously mentioned case. In the embodiment illustrated in FIG. 2, the surrounding capsule 21, 22 is dispensed with. Instead, the container 10 and the lid 18 are provided with flanges 25 and 26, respectively. After placing the fuel rods 11 in the container and filling this with copper powder 17, the flanges 25 and 26 are joined together by welding or cold pressing to form a gas-tight joint 27. The lid 18 is provided with a tube 28 of copper which is sealed after evacuation of the container and its gas-tight lid. After sealing of the tube 28, the closed container is subjected to isostatic pressing in two stages in either of the manners described for the sealed capsule in accordance with FIG. 1. |
062298712 | abstract | A chuck for holding a reflective reticle where the chuck has an insulator block with a non-planer surface contoured to cause distortion correction of EUV radiation is provided. Upon being placed on the chuck, a thin, pliable reflective reticle will conform to the contour of the chuck's non-planer surface. When employed in a scanning photolithography system, distortion in the scanned direction is corrected. |
claims | 1. A machine ( 1 ) for producing energy by nuclear fusion reactions, characterized by comprising a reaction chamber ( 2 ); a target ( 5 ) housed inside said reaction chamber ( 2 ); a positive deuterium ion source ( 3 ) communicating with said reaction chamber ( 2 ); and a pumping assembly ( 4 ) communicating with said reaction chamber ( 2 ) to maintain a vacuum inside the reaction chamber ( 2 ); said target ( 5 ) comprising an active element ( 31 ) made of a material having a crystal lattice and deuterium atoms in its crystal lattice; and said positive deuterium ion source ( 3 ) feeding a flux of positive deuterium ions into the reaction chamber ( 2 ), so that said flux of positive deuterium ions strikes said active element ( 31 ) of the target ( 5 ) to produce nuclear fusion reactions between the incident positive deuterium ions and the atoms constituting the target ( 5 ) itself, said target ( 5 ) comprising an outer casing ( 30 ), which is substantially cup-shaped with a cavity ( 30 a ) facing said focusing device ( 13 ); said active element ( 31 ) being housed inside said outer casing ( 30 ) and being substantially cup-shaped with a cavity ( 31 a ) facing a focusing device ( 13 ); said active element ( 31 ) being struck by a concentrated beam of positive deuterium ions. 2. A machine as claimed in claim 1 , characterized in that said reaction chamber ( 2 ) comprises accelerating means ( 10 ) for conveying said flux of positive deuterium ions against said target ( 5 ) and accelerating the positive deuterium ions. claim 1 3. A machine as claimed in claim 2 , characterized in that said accelerating means ( 10 ) comprise at least two electrodes ( 5 , 12 ) housed inside the reaction chamber ( 2 ); and an electric energy source ( 11 ) for maintaining an in any way variable difference in electric potential between the two electrodes ( 5 , 12 ); one of said two electrodes ( 5 , 12 ) being defined by said target ( 5 ). claim 2 4. A machine as claimed in claim 1 characterized in that said reaction chamber ( 2 ) comprises a focusing device ( 13 ) through which said flux of positive deuterium ions travels prior to striking said target ( 5 ); said focusing device ( 13 ) focusing said flux of positive deuterium ions to form a concentrated beam of positive deuterium ions. claim 1 5. A machine as claimed in claim 1 , characterized in that said outer casing ( 30 ) is made of metal material. claim 1 6. A machine as claimed in claim 5 , characterized in that said outer casing ( 30 ) is defined by a stack of alternating washers ( 32 ) made of different metal materials and held together by gripping means. claim 5 7. A machine as claimed in claim 1 , characterized in that said active element ( 31 ) is defined by an aggregate comprising metal salts. claim 1 8. A machine as claimed in claim 7 , characterized in that some of the metal salts belong to the class of sulfates comprising ferrous sulfate, nickel sulfate, titanium sulfate and potassium sulfate. claim 7 9. A machine as claimed in claim 7 , characterized in that the metal salts are selected from copper sulfate and lithium sulfate; said sulfate being hydrated with heavy water. claim 7 10. A machine as claimed in claim 1 , characterized in that said material of said active element ( 31 ) is capable, when stricken by a deuterium ion flux, of absorbing positive deuterium ions and starting an endothermic reaction with said positive deuterium ions. claim 1 11. A machine as claimed in claim 1 , characterized in that said positive deuterium ion source ( 3 ) comprises a tank ( 15 ) containing deuterium atoms in gaseous form for supply to said reaction chamber ( 2 ); and an ionizing unit ( 16 ) interposed between said tank ( 15 ) and said reaction chamber ( 2 ) to so ionize the deuterium atoms as to supply said reaction chamber ( 2 ) with a flux of positive deuterium ions. claim 1 |
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claims | 1. A system for controlling particles using projected light, the system comprising:a particle system configured to provide a plurality of particles;an optical source configured to generate a beam of light with a frequency shifted from an atomic resonance of the plurality of particles; anda beam filter positioned between the particle system and plurality of particles, and comprising a first mask, a first lens, a second mask, and a second lens,wherein the optical source, beam filter, and particle system are arranged such that the beam of light from the optical source passes through the beam filter, and is projected on the plurality of particles to form an optical pattern that controls the positions of the particles in space. 2. The system of claim 1, wherein the first mask is positioned a first focal length away from the first lens, and the second mask is positioned a first focal length away from the first lens and a second focal length away from the second lens. 3. The system of claim 2, wherein the first mask, the first lens, the second mask, and the second lens are arranged such that the beam of light passes sequentially therethrough. 4. The system of claim 1, wherein the first mask comprises a reflecting plane formed using a substrate coated with a reflective layer. 5. The system of claim 4, wherein the reflective layer comprises at least one transmitting region producing at least one bright region in the optical pattern. 6. The system of claim 4, wherein the reflective layer comprises at least one reflecting region producing at least one dark region in the optical pattern. 7. The system of claim 1, wherein the beam filter further comprises a third mask positioned between the first mask and the first lens. 8. The system of claim 7, wherein the third mask is a phase scrambling mask having phase scrambling regions configured to transmit and impart a phase shift to light passing therethrough. 9. The system of claim 8, wherein phase shifts imparted by different phase scrambling regions are different, and distributed randomly across the phase scrambling mask. 10. The system of claim 1, wherein the first mask comprises a plurality of apertures in a one-dimensional (1D) or two-dimensional (2D) array. 11. The system of claim 1, where the plurality of particles comprises neutral atoms. 12. The system of claim 1, wherein the beam of light has a frequency shifted from the atomic resonance to achieve a blue detuning or a red detuning. 13. The system of claim 1, wherein the beam filter is further configured to transform a Gaussian beam or a near-Gaussian beam into a beam with a uniform intensity profile. 14. A method for controlling particles using projected light, the method comprising:generating a beam of light using an optical source;directing the beam of light to a beam filter comprising a first mask, a first lens, a second mask, and a second lens;forming an optical pattern using the beam filter; andprojecting the optical pattern on a plurality of particles to control their locations in space. 15. The method of claim 14, wherein the first mask is positioned a first focal length away from the first lens, and the second mask is positioned a first focal length away from the first lens and a second focal length away from the second lens. 16. The method of claim 15, wherein the first mask, the first lens, the second mask, and the second lens are arranged such that the beam of light passes sequentially therethrough. 17. The method of claim 14, wherein the first mask comprises a reflecting plane formed using a substrate coated with a reflective layer. 18. The method of claim 17, wherein the reflective layer comprises at least one transmitting region producing at least one bright region in the optical pattern. 19. The method of claim 17, wherein the reflective layer comprises at least one reflecting region producing at least one dark region in the optical pattern. 20. The method of claim 14, wherein the beam filter further comprises a third mask positioned between the first mask and the first lens. 21. The method of claim 20, wherein the third mask is a phase scrambling mask having phase scrambling regions configured to transmit and impart a phase shift to light passing therethrough. |
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047598980 | claims | 1. A liquid metal-cooled fast neutron nuclear reactor comprising: a vessel, a horizontal slab closing said vessel; a core cover plug suspended from said slab and having a vertically axed external ferrule surrounding control rod guidance sleeves and also vertically axed liquid metal sampling tubes; said ferrule supporting a horizontal core cover plate traversed by said sleeves and tubes; said core cover plug also comprising: at least one grid located below the core cover plate; each grid having a horizontal, annular, peripheral plate connected in fixed manner to the core cover plate and also having a plurality of horizontal modular plates separate from said peripheral plate, each of said modular plates being fixed to at least one of the sleeves and/or tubes, said modular plates and said peripheral plate defining therebetween passages for the liquid metal, wherein the modular plates of the same grid are vertically downwardly displaced on moving away from the said ferrule axis. 2. A liquid metal-cooled fast neutron nuclear reactor comprising: a vessel, a horizontal slab closing said vessel; a core cover plug suspended from said slab and having a vertically axed external ferrule surrounding control rod guidance sleeves and also vertically axed liquid metal sampling tubes; said ferrule supporting a horizontal core cover plate traversed by said sleeves and tubes; said core cover plug also comprising: at least one grid located below the core cover plate; each grid having a horizontal, annular, peripheral plate connected in fixed manner to the core cover plate and also having a plurality of horizontal modular plates separate from said peripheral plate, each of said modular plates being fixed to at least one of the sleeves and/or tubes, said modular plates and said peripheral plate defining therebetween passages for the liquid metal, the modular plates of the same grid being vertically upwardly displaced on moving away from said ferrule axis. 3. A reactor according to claim 2, wherein edges of adjacent modular plates and the peripheral plate of the same grid form therebetween clearances in a horizontal direction. 4. A reactor according to claim 2, wherein edges of adjacent modular plates and the peripheral plate of the same grid are vertically aligned. 5. A reactor according to claim 2, wherein edges of adjacent modular plates and the peripheral plate of the same grid partly overlap in a horizontal direction. 6. A reactor according to claim 2, wherein said ferrule is fixed to the peripheral plate and has perforations below the core cover plate. 7. A reactor according to claim 2, wherein the lower end of the ferrule is fixed to the core cover plate, the peripheral plate being suspended on the core cover plate by tie bolts. 8. A reactor according to claim 7, wherein the external diameters of the core cover plate and the peripheral plate slightly exceed the external diameter of the ferrule, a break being provided on the periphery of the peripheral plate for the passage of handling means. 9. A reactor according to claim 2, wherein the core cover plug comprises a deflecting grid located in the vicinity of lower ends of the sleeves and tubes and a grid forming a heat shield located between the deflecting grid and the core cover plate. |
047643331 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT With reference to FIGS. 1 to 3, a flask door 1 is mounted in a flask base 2 which, in turn, is secured to an end of a flask body (not shown). In the position illustrated in FIGS. 2 and 3 the door 1 closes an opening in the base 2 and through which contents of the flask can be discharged. The door 1 comprises complementary upper and lower wedge-shaped members 4 and 5 respectively which together form a door of substantially uniform thickness which is received in the flask base 2. The lower member 5 is provided with wheels or rollers 6 which run on rails 7 located on inwardly directed flanges on side walls of the flask base. The upper member 4 carries a continuous seal which cooperates with a sealing ring 9 which is secured in the base and about the opening 3. To assist sliding movement between the wedge-shaped members 4 and 5, bearing strips 10 are provided in the upper surface of the lower member 5 to engage the lower surface of the upper member 4. The flask door 1 cooperates with a gate 11 (FIG. 4) which is mounted in a gate housing 12 (FIGS. 5 and 6). The flask base 2 containing the flask door 1 is received in the gate housing 12 and the door 1 mechanically interlocks with the gate 11 as a result of upstanding dowels 13, 14 on the gate which are received in mating dowel holes 15, 16 in the door. The gate 11 comprises two, co-planar, inter-leaved portions 17, 18 which are mounted on wheels 19 to run on rails 20 in the gate housing. Dowels 13 on the gate portion 17 cooperate with dowel holes 15 in the end of the upper wedge-shaped member 4 of the door 1. Dowels 14 on the gate portion 18 cooperate with dowel holes 16 in the lower wedge-shaped member 5 of the door 1. The inter-leaved gate portions 17, 18 are urged apart by means of two spring loaded separation mechanisms 21, only one of which is shown in FIG. 4, positioned symmetrically at opposite sides of the centre line of the gate. Each mechanism 21 comprises a compression spring 22 contained within a housing 23 in the gate portion 18, the spring 22 acting on a plunger assembly 24 slidable in the portion 18 and secured at its end remote from the spring to the gate portion 17. Cooperating stops 25, 26 on each side of the respective gate portions 17, 18 limit the extent of the separation effected by the spring loaded separation mechanisms 21. A drive mechanism for moving the gate and door assembly, the gate being coupled to the door by the dowels 13, 14 is shown in FIGS. 5 and 6. The drive mechanism comprises a drive motor 27 which is coupled through a chain 28 and sprocket 29 to the input shaft 30 of a bevel gear unit 31. The motor 27 can be provided with a manually operated handle 32 to enable the gate and door assembly to be moved in the event of a power failure. A lead screw 33 extends at 90.degree. to the input shaft 30 from the bevel gear unit 31 to a journal 34 in the housing 12. The lead screw 33 imparts linear motion to a crosshead 35 supporting two pinions 36. The pinions 36 mesh with racks 37 fixed in the base of the gate housing 12 and with racks 38 located in channels 39 (FIG. 4) in the base of gate portion 18. The fixed racks 37 impart rotation to the pinions 36 which in turn, through the racks 38, effects displacement of the gate and door assembly. The arrangement is such that the linear displacement of the gate and door assembly is twice that of the pinions along the racks (compare FIG. 5 and FIG. 9). Starting from a position at which the door 1 is in sealing engagement with the base 2 by virtue of the seal 8 cooperating with the sealing ring 9, the assembly operates in the following manner. Actuation of the drive motor 27 effects linear displacement of the gate portion 18 and the separation of the gate portions 17 and 18. The spring loaded mechanisms 21, which can be adjustable, operate to urge the gate portion 17 away from the gate portion 18. Separation of the two gate portions proceeds until the stops 26 on the portion 18 abut against the stops 25 on the portion 17. Thereafter continued operation of the drive motor 27 causes the two gate portions to move together as a unit. As the wedge-shaped door members 4 and 5 are fixedly secured to the gate portions 17 and 18 respectively by means of the dowels 13, 14 it follows that the door members move with the gate portions. The initial movement of the gate portion 18 effects similar movement of the lower wedge-shaped door member 5 to cause separation of the door members 4 and 5 at their inclined surfaces. Over this initial movement, the extent of which is determined by the positions of the stops 25 and 26, the gate portion 17 and the upper wedge-shaped door member 4 are stationary. The resulting gap created between the inclined surfaces of the now separated door members allows the upper door member 4 to fall vertically away from the flask base 2 thus breaking the seal therebetween. This initial movement therefore avoids relative sliding movement between the seal 8 and the sealing ring 9 which could result in scuffing and tearing of the seals. Thereafter the door members 4 and 5 move with the gate portions 17 and 18 into a fully open position to permit unimpeded access through the gate housing 12 into a flask located on the housing. FIGS. 7 to 9 depict the sequence of movements of the door and gate assembly between a fully closed position (FIG.7); an intermediate position at which the seals break (FIG. 8); and a fully open position (FIG. 9). To close and seal a flask the drive motor is reversed to return the door and gate assembly. The gate portions 17 and 18 and the respective door members 4 and 5 are maintained separated by the spring loaded mechanisms 21 until the leading end of the upper door member 4 abuts against the stop face of the base 2. At the same time the leading end of the gate portion 17 abuts against face 40 of the gate housing 12. Continued movement of the gate portion 18 overcomes the spring loaded mechanisms 21 to close the gap between the portions 17 and 18 and the portions 18 finally abuts against the gate portion 17. The lower door member 5 moves with the gate portion 18 and in so doing it displaces the upper door member 4 vertically upwards to effectively bring the seal 8 into sealingly engagement with the sealing ring 9 in the flask base 2. |
abstract | Methods and apparatus are provided for scanning a charged particle beam. The apparatus includes scan elements and a scan signal generator for generating scan signals for scanning the charged particle beam in a scan pattern having a scan origin. In one embodiment, the apparatus includes a position controller for positioning the scan elements based on a parameter of the charged particle beam, such as energy. The scan elements may be positioned to achieve a fixed position of the scan origin for different beam energies. In another embodiment, the apparatus includes first and second sets of scan elements and a scan signal controller for controlling the scan signals supplied to the sets of scan elements based on a parameter of the charged particle beam, such as energy. The scan signal controller may control the ratio of the scan signals applied to the sets of scan elements, or may deenergize a set of scan elements, to minimize space charge forces on the charged particle beam that may reduce beam transmission through the apparatus. |
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046735453 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT With reference to FIG. 1, a typical fuel rod assembly 30 includes fuel rods 22, a plurality of fuel rod spacers 46, and a top cap 47. A number of anti-baffle jetting or anti-fretting clips 31 are shown installed at various positions on the rods 22 of the fuel assembly 30 for substantially preventing fretting or vibrational damage to the rods 22. FIG. 2 is a cross-sectioanl view along lines 2--2 of fuel assembly 30, showing a top view of a clip 31 securing together a number of fuel rods 22, and an inert rod 40. The inert rod 40 is shown with a tab 41 welded along its length, which tab 41 abutts the fuel assembly or fuel rod spacers 42 and 43. FIG. 3 is a close-up pictorial view of an enlarged portion of a clip 31 as installed to secure a number of fuel rods 22 and inert rod 40 together. For greater detail of the use and design of the clips 31, reference is made to previously mentioned co-pending U.S. Ser. No. 487,907, filed Apr. 25, 1983, and assigned to the same assignee as the present invention. In the typical nuclear electrical generating power station a large tank of water is utilized to enclose spent or irradiated nuclear fuel rod assemblies. Such an installation is shown in FIG. 4, and includes a pictorial of the apparatus of the subject invention. As shown, a deep tank 50 is substantially filled with water 52, for safely storing a number of irradiated fuel assemblies 30 in a spent or irradiated fuel assembly rack 54 located in the bottom of tank 50. A long grapple 56 extending from an overhead crane (not shown), is shown connected via a grapple rod 58 and grapple stop plate assembly 60 to one of the fuel rod assemblies 30 partially inserted in a cell 138 of the fuel rack 54. The combination of the overhead crane and long grapple 56 is operable for moving fuel assemblies 30 about the tank 50, and either into or out of the spent fuel rack 54. A working platform 62 is mounted above the tank 50 and protrudes partly over the water 52. A safety railing 64 surrounds the perimeter of the working platform 62 to help prevent an operator 66 from falling off the platform 62. An operator and tool support bracket is formed by a beam 68 secured to the working platform 62. The beam 68 has a cantilever portion projecting from the working platform 62 over the water 52 of the tank 50 as shown. A "fresh fuel elevator assembly" 70 is mounted in the tank as shown, and includes a motor 72 rigidly held above the surface of the water 52, as shown. As shown by the arrow path 74, the long grapple 56 is used for moving fuel assemblies 30 between the fuel rack 54 and the fresh fuel elevator 70. An irradiated fuel assembly 30, as depicted by the arrow 76 is shown just after being moved from the fuel rack 54 to mounting upon the fresh fuel elevator assembly 70. Clips 31 must be removed from this fuel assembly 30. PG,9 The present apparatus for permitting the removal of clips 31 from a fuel assembly 30 includes the operator support bracket 68, a disposal basket 78 suspended by a cable 80 having a hook 82 at one end for coupling to an eyelet 84 rigidly connected to the end of the cantilever beam comprising the operator support bracket 68. The other end of the cable 80 is secured to another hook 86 for coupling to a suspension ring 88 secured to the clip disposal basket 78. The disposal basket 78 also includes an access window 90 for receiving clips 31 removed from fuel assembly 30. A television monitor 92 is secured to the top of the railing 62 in an appropriate position for viewing by an operator 66 positioned as shown. The operator 66 is shown in the act of positioning a clip removal tool assembly 94. The clip removal tool assembly 94 includes a tool support pole or tubing 96 secured at an upper portion thereof to a carriage assembly 98, thereby establishing the height of the tubing 96 protruding above the operator support bracket 68, and corresponding the length of the tubing 96 protruding below the operator support bracket 68 to a predetermined depth (about 40 feet, for example) within the water 52 of tank 50. A spoked wheel 100 is rigidly connected to an uppermost portion of the tubing 96. The lowermost end of the support tubing 96 is rigidly connected to the top of a frame 102 of a clip removal tool 104, thereby establishing the location of the clip removal tool 104 at a predetermined depth and area within tank 50. As shown, the frame 102 includes upright brackets 106 for rigidly mounting a television camera 108 at a predetermined angle in the tool 104 for viewing the operating area of a jaw assembly 110. The jaw assembly 110 is rigidly mounted to the bottom of a generally U-shaped frame 112. The frame 112 is rigidly attached to the upper frame 102 via rigid mounting straps 114, as shown. A portion of a hydraulic cylinder 116 is shown mounted between the rear of the jaw assembly 110 and a cross bracket 118 rigidly mounted between the inside faces of the mounting straps 114. Also shown is a right-hand portion of a mirror system 120, which includes an air or hydraulic cylinder 122 mounted to a bottom surface of a jaw assembly 110 for providing selective positioning of the right-hand mirror 124. An identical mirror system is included on the left-hand side of the jaw assembly 110, but is not shown in the view given in FIG. 4. Note that the fresh fuel elevator 70, although not part of the present apparatus, is used in conjunction with the operation of the present apparatus in order to accomplish the removal of the clips. The fresh fuel elevator 70 includes a carrier 126 which is moved up and down along a central post 128, and two guide posts 130 and 132, as shown. The motor 72 is used to rotate a pulley 134 for winding and unwinding a cable 136 attached to the carrier 126 for positioning the carrier 126 at a desired vertical location in the tank. As previously mentioned, the fresh fuel elevator 70 is typically for use in facilitating handling of fuel assemblies 30 within the water tank 50. With reference to FIGS. 4 and 5, a general description of the operation of the present apparatus will now be given. Typically, an irradiated fuel assembly 30 is stored in the spent fuel rack 54 until such time it si to be either disposed of or installed in a reactor core. An irradiated fuel assembly 30 removed from a peripheral location in a reactor core (not shown), may be temporarily stored in the spent fuel rack 54 until such time that the clips 31 are to be removed therefrom, for permitting the irradiated fuel bundle 30 to be restrained in an interior location of the reactor core. Assume that as shown in FIG. 4, the grapple 56, shown attached at an upper end to the hook and cable assembly 140 of an overhead crane (not shown), and connected at its other end to an irradiated fuel assembly 30 previously stored in a cell 138 of fuel rack 54, was operated to lift the fuel assembly from the fuel rack 54 to a position in the fresh fuel elevator carrier 126 (see arrow 76) as shown. The grapple stop plate 60 serves to hold the fuel assembly 30 in place in carrier 126 via the coupling rod 142. The fuel assembly 30 was previously dropped through a hole in the carrier now covered by the grapple stop plate 60 until the stop plate 60 seats on the carrier plate 126 with the fuel assembly 30 appropriately oriented to the clip removal tool assembly 104, for removal of certain ones of the clips 31 from the assembly 30. By observing the positioning of the jaw assembly 110 to the fuel assembly 30, the operator 66 first operates the fresh fuel elevator 70 controls (not shown) to position a clip 31 to be removed in the same horizontal plane as the jaw assembly 110. The operator 66 next proceeds to grasp the spoked wheel 100 to both turn the spoked wheel 100 and push the carriage assembly 98 forward and backwards as required along a slot 99 in the operator support bracket 68, for positioning the jaws 110 immediately opposite and typically within an inch of the clip 31 to be removed. This is accomplished by observing the view of the area about the jaw assembly 110 on the television monitor 92, and also by operating the appropriate controls (not shown) for operating the mirror system 120 (and the mirro system on the opposite side of the jaw assembly 110) to obtainthe best view of the area about the jaw assembly 110. The operator 66 next pushes the spoked wheel 100 in a direction away from the fuel assembly 30, thereby causing the carriage assembly 98 and tubing 96 to "tilt" in a direction forcing the jaw assembly 110 into contact with portions of fuel rods 22. When properly positioned, the upper and lower jaws 144, 146, respectively, are in the open position about a clip 31 as shown in FIG. 5A. As will be described in greater detail later, as also shown in the view, an upper support plate 148, and a lower support plate 150, are abutted against portions of the fuel rods 22 (shown in phantom in view A of FIG. 5) above and below the clip 31 to be removed. The hydraulic actuator 116 is then operated for causing the jaws 144 and 146 to close on the top and bottom edges of the clip 31, as shown in FIG. 5B. Continued operation of the hydraulic actuator 116 causes the jaws 144 and 146 to move away from the fuel rods 22, thereby pulling the clip 31 away from engagement with the fuel rods 22, as shown in FIG. 5C. The operator 66 next proceeds to pull the spoked wheel 100 toward the fuel assembly to "tilt" the jaw assembly 110 away from the fuel assembly 30. The spoked wheel 100 is then turned in a clockwise direction, in this example, to position the front of the jaw assembly 110 within the window 90 of the disposal basket 78. The last step in the quick removal operation, is for the operator 66 to reverse the previous operation of the hydraulic actuator 116 for moving the jaws 144 and 146 forward towards the front of the jaw assembly 110 to their open position, as shown in FIG. 5D. The operator 66 next operates a pair of ejector cylinders (to be described later) for causing a pair of ejector plungers 148 (shown in phanthom in FIG. 5D) to move forward for pushing the clip 31 out of the jaw assembly 110 and into the basket 78. the next clip 31 to be removed from the fuel assembly 30 is so removed by repeating the previously described sequence of operation, including operation of the fresh fuel elevator 70, if necessary. Note that as the jaws 144, 146 are moved back and forth between the upper and lower support plates 148,150, cam-like surfaces 145 and 147 of jaws 144,146, respectively, interact with cam surfaces 149 and 150 on the inside faces of the upper and lower support plates 148,150, respectively, for opening and closing the jaws 144,146, as shown. Details of the carriage assembly 98 will now be described with reference to FIGS. 6 through 10. The carriage assembly 98 includes a base plate 154, an upper plate 156, cap screws 158, tube clamp 160, carriage travel control bar 162, pivot bar spacers 164, clamping plates 166, a tube protection cap 168, machine screws 170, auto track ball transfers 172 tack welded in place, bearings 174, and jam nuts 176. These components are assembled as shown in FIGS. 6 through 10. Note that two carriage control bars 162 may be secured on either side of the base plate 154 across the slotway 99 in order to limit the movement of the carriage assembly 98 in the forward and backward positional extremes. The support tubing 96 is rigidly secured at an appropriate portion thereof, via the tubing clamp 160. As shown, the tubing clamp 160 includes two halves 178 and 180. The half tubing clamp portion 178 includes a pair of threaded holes for receiving cap screws 158 inserted in holes in the other half portion 180, for securing the tubing clamp 160 to the tubing 96, as shown. As shown in FIG. 4, a hole 182 is provided in the upper portion of the tubing 96 near the working platform 62 for receiving control cables 184 (note that the cables include electrical wires and hydraulic or pressurized air lines) for connection to the TV camera 108, the hydraulic actuator 116, and the mirror systems including system 120. A pictorial view of the carriage mechanism 98 is shown in FIG. 11, and also includes a pictorial view of a typical spoked wheel 100 connected to tubing 96 shown in ghost or phantom view. In this example, the spoked wheel 100 is shown to include two sections which are clamped about tubing 96, and also shown protruding from each half section of the spoked wheel 100 are spokes or handles 190. By turning the spoked wheel 100 clockwise or counterclockwise the tubing 96 is easily rotated via the corresponding rotation of the upper carriage plate 156 upon the ball-like bearings 172 mounted on the top of the carriage base plate 154. Also, by either pushing or pulling on the handles 190, the tubing 96 is easily moved forward or backwards within a range along the slotway 99 in the operator support bracket 68, via the interaction between the ball-like bearings 172 on the bottom of the carriage base plate 154 rolling along the top surface of the operator support bracket 68, and the interaction between the roller-like bearings 174 rolling along the inside faces of the operator support bracket 68. Design details of the jaw assembly 110 follow with reference to FIGS. 12 through 15. As shown, the jaw assembly 110 includes an upper support plate 148 having finger-like projections 182 along the width of its front or leading edge for engaging juxtaposed fuel rods 22 (five in this example) about portions of the fuel rods 22 above a clip 31 to be removed (see FIG. 12), thereby supporting the fuel rods 22 above a clip 31 to be removed. Further included are a pivot bracket 184 for connecting hydraulic or air cylinder 116 to the cross bracket 118, cap screw 186, machine screws 188, a right-hand side plate 190, a left-hand side plate 192, a set collar 194, an upper support plate 148, a jaw actuator cam 149, a lower support plate 150, a jaw actuator cam 151, an upper clip removal jaw 144, a lower clip removal jaw 146, jaw support pins 196, a ball bushing 198, a jaw support block 200, a linear bearing support block 202, a retaining ring 204, an actuator shaft 206, a hex nut 208, an actuator handle 210, a straight linkage 212, a bent linkage 214, a plunger 216 of cylinder 116 and secondary actuator shaft 218, all assembled as shown. As shown in FIGS. 14 and 15, the clip removal jaws 144 and 146 are identical. The jaws 144 and 146 include finger-like projections 220 (four such fingers 220 are shown in this example), with semi-circular cutouts 222 provided between the fingers 220 for permitting the fingers 220 to slip between three juxtaposed fuel rods 222 to a depth where the notches 224 of each finger 220 can be placed over an edge of a clip 31 to be removed from a fuel assembly 30. As previously described, the jaws 144, 146 include identical cam surfaces 145 and 147, contoured in this example to have angles alpha (.alpha.) of 30.degree., beta (.beta.) of 45.degree., gamma (.gamma.) of 7.degree., and rho (.rho.) of 10.degree. (the latter angle being associated with the notch 224). A hole 226 is provided for receiving pins 196 for pinning the jaws 144, 146 to the jaw support block 200 (see FIG. 20). In FIG. 16, a pictorial view is shown of the jaw support block 200. Mounted on the back or rear face of the jaw support block 200 are two relatively small air cylinders 228, in this example, screwed into threaded holes of the former. As will be later described in more detail, actuation of the cylinders 228 causes respective plungers to rapidy move outward from the front face of the jaw block 200 for pushing a clip 31 out of the jaw assembly 110 (see FIG. 21). The jaw block 200 also includes a threaded hole 232 for receiving the threaded end of the actuator shaft 206 (see FIG. 13). Details of the right and left-hand mirror assemblies 120, 232, respectively, are shown in FIGS. 17 and 18. As shown, the mirror assemblies 120 and 232 are identical, and each include in this example a circular mirror 234, a mirror extension bolt 236, a hex lock nut 238, a support arm end fitting 240, a mirror support arm extension 242, Allan Head set screws 244, cap screws 246, hydraulic or pneumatic activator cylinders 248, connecting bolts 250, flat washers 252, swivel busings 254, pivot bolts 256, and lock nuts 258, assembled as showm. Individual actuation of either one of the actuator cylinders 248 permit an operator 66 to change the angular orientation of either of the mirrors 234 relative to the front of the jaw assembly 110, for changing the field of view reflected from the mirrors 234 to the television camera 108. FIG. 19 shows a typical field of view as viewed from the television monitor 92 (the field of view shown is that actually viewed from the camera 108). A more detailed description of the operation of the present invention, particularly the jaw assembly 110, will now be given with reference to FIGS. 4, 13, and 20, for removing a clip 31 from a fuel assembly 30. The operator 66, as previously described, operates the fuel elevator 70, in conjunction with turning, pushing and pulling the spoked wheel 100 and moving the carriage assembly 98 forward or backward, as required, to position the tool 104 with its jaw assembly 110 positioned about a clip 31 as shown in FIG. 20. As shown in FIG. 20, the jaws 144 and 146 are open with their notches 224 immediately over the top and bottom edges of the clip 31. The fingers 182 of the upper support plate 148 support the fuel rods 2 at portions above the clip 31, whereas the leading edge of the lower support plate 150 supports the fuel rods 22 below the clip 31 to be removed. The operator 66 may periodically activate either or both of the cylinders 248 for repositioning the mirrors 234 for improving his view of the area about the clip 31 to be removed. With the jaws 144 and 146 in the open position, the actuator cylinder 116 has its plunger 216 positioned as shown in FIG. 13, whereby the actuator shaft is positioning the jaw support block 20 in its forwardmost position, where the interaction between the jaws 144 and 146, with the cams 149 and 151, respectively, holding the jaws 144, 146 in their open position. The operator 66 next activates the cylinder 116 for causing the plunger 216 thereof to move towards the positions shown in phantom in FIG. 13, causing the jaw block 200 to move rearward, whereby the interaction between the jaws 144 and 146, and cams 149 and 151, respectively, causes the jaws 144, 146 to very rapidly close for tightly grasping the top and bottome dges of the clip 31 in the jaw notches 224 (occurs upon only slight rearward movement of the jaw block 200). As the plunger 216 continues to move toward the phantom position shown in FIG. 13, the jaw block 200 continues to move rearward pulling the clip 31 away from the fuel assembly 30, whereby when the plunger 216 reaches the phantom position of FIG. 13, the clip 31 will be completely removed from the fuel assembly 30. Once a clip has been removed from a fuel assembly 30, the operator 66 must next proceed to deposit the removed clip 31 into the disposal basket 78 (see FIG. 4). To accomplish this, the operator 66 observes the TV monitor 92 and manipulates the spoked wheel 100 for positioning the front of the jaw assembly 110 into the window 90 of disposal basket 78. The operator 66 then activates the activator cylinder 116 for moving the jaw support block 200 forward to cause the jaws 144 and 146 to open. Next, the ejector cylinders 228 are activated for causing their respective plungers 230 to move rapidly outward from holes 242 and 244 within a recess 240 of the jaw block 200 (see FIG. 21), causing the removed clip 31 to be ejected into the disposal basket 78. In this manner, the apparatus of the present invention permits irradiated clips 31 to be safely removed from an irradiated fuel assembly 30. While the present invention has been described in connection with the preferred embodiments thereof, it should be understood that there may be other obvious modifications or embodiments of the present invention which fall within the spirit and scope of the invention as defined by the appended claims. |
039717321 | abstract | An apparatus for fixing radioactive or toxic waste has an extruder including a mixing mechanism for intermingling and advancing the waste and a carrier material introduced into the extruder. The extruder has a heating zone with which there communicates a vapor outlet device having an observation window. Within the vapor outlet device there is disposed an arrangement for cleaning the window and an arrangement for removing deposits from those locations of the vapor outlet device that are adjacent the mixing mechanism. The condenser is coupled to a distillate accumulator with the interposition of two alternatingly operating filters for removing particles from the condensate obtained from the condenser. To the outlet of the extruder there is coupled a loading device in which containers are successively filled with the material discharged by the extruder. The loading device includes an interrupter bowl which receives the material discharged by the extruder during an exchange of an empty container for a filled container below the extruder outlet. |
description | The present application claims priority to U.S. Provisional Application No. 61/219,655, filed Jun. 23, 2009, the entirety of which is incorporated herein by reference. The present invention relates to a circumferential sampling tool. One method of assessing the useful life of pressure tubes in nuclear reactors, such as a CANDU reactor, requires the periodic removal of a tube. Samples are cut from the removed tube and analyzed for deuterium content. The deuterium concentration is then used as a measure of the useful life of the remaining pressure tubes. This approach is very costly because of the long shutdown period required to remove and replace a pressure tube. Attempting to provide in-situ sampling (without pressure tube removal) presents numerous difficulties. Obtaining a useful sample is made difficult by the hard oxidized surface, and the need to obtain sample material from beneath the surface layer. To preserve the structural integrity of the tube and avoid detrimental residual stress, the sampling depth must be controlled and the sampled region must be left with smooth changes in geometry in all axes. Furthermore, the technique used for removing the surface material or sample must not involve excessive heating, as this affects the results of the subsequent analysis. Another difficulty is the recovery of the sample for analysis and preventing particles from being left in the pressure tube. U.S. Pat. No. 4,925,621, issued May 15, 1990, the entirety of which is incorporated herein by reference, discloses a sampling tool useful for pressure tube sampling which addresses the above difficulties. The disclosed sampling tool permits in situ testing in that pressure tube removal is unnecessary. The sampling tool comprises two cutters and means for capturing the removed material. By moving both cutters axially in the pressure tube, one cutter removes the surface oxide layer, and the second cutter removes a sample for analysis. The cutters and cutting operation are designed to avoid damaging the integrity of the pressure tube to allow it to remain in service. Although the above-described sampling tool addresses the above difficulties, it proves impractical to obtain samples in some portions of the pressure tube. For example, as seen in FIG. 1, in a CANDU type fuel channel, the pressure tube 10 is joined to an end fitting (not shown) using a rolled joint 12. The above-described sampling tool makes obtaining useful samples in the rolled joint area difficult due to the high axial gradient of hydrogen/deuterium concentration and the circumferential ripples 14 in the rolled joint area. The conference paper presented at the 5th International CANDU Maintenance Conference in November 2000 which is entitled “Advanced Pressure Tube Sampling Tools” and is authored by K. Wittich and J. King also discloses sampling tools. The conference paper presented at the 7th International CANDU Maintenance Conference in November 2005 which is entitled “Innovation in Pressure Tube Rolled Joint Sampling (Circumferential Sampling Tool Technology)” and is authored by B. Guler, J. King, and R. Wray also discloses sampling tools. Both papers are published by the Canadian Nuclear Society. Therefore, there is a need for a sampling tool that addresses at least some of the above-identified difficulties and at least some of the inconveniences present in the prior art. It is an object of the present invention to provide a sampling tool that has at least two cutters that move circumferentially along a portion of an interior wall of a tube. One cutter removes a portion of the interior wall of the tube, and the second cutter removes a sample from the interior wall of the tube from a location in the tube revealed by removing the portion of the interior wall of the tube. In one aspect, a circumferential sampling tool for obtaining a sample from an interior wall of a tube has a cylindrical body having a central axis, an aperture in the cylindrical body, and a shaft disposed in the cylindrical body along the central axis. A first cutter is operatively connected to the shaft for rotation therewith. The first cutter is movable radially between a retracted position where the first cutter is disposed inside the cylindrical body at a first distance from the central axis and an extended position where the first cutter extends at least in part through the aperture at a second distance from the central axis. The second distance is greater than the first distance. A first actuator is operatively connected to the first cutter for moving the first cutter between the retracted position and the extended position as the shaft rotates. The first actuator mechanically biases the first cutter toward the retracted position. A second cutter is operatively connected to the shaft for rotation therewith and is disposed at an angle to first cutter. The second cutter is movable radially between a retracted position where the second cutter is disposed inside the cylindrical body at a third distance from the central axis and an extended position where the second cutter extends at least in part through the aperture at a fourth distance from the central axis. The fourth distance is greater than the third distance. The fourth distance is greater than the second distance. A second actuator is operatively connected to the second cutter for moving the second cutter between the retracted position and the extended position as the shaft rotates. The second actuator mechanically biases the second cutter toward the retracted position. The second cutter is in the retracted position when the first cutter is in the extended position. The first cutter is in the retracted position when the second cutter is in the extended position. Rotating the shaft causes the first cutter to move to the extended position thereby cutting a portion of the interior wall of the tube and then causes the second cutter to move to the extended position thereby cutting the sample from the interior wall of the tube from a location in the tube revealed by cutting the portion of the interior wall of the tube. In an additional aspect, the first actuator has a spring mechanically biasing the first cutter toward the retracted position. The second actuator has a spring mechanically biasing the second cutter toward the retracted position. In a further aspect, a ramp is disposed inside the cylindrical body along a circumferential portion thereof. The ramp is disposed opposite the aperture. The first actuator also has a first roller. The first roller causes the first cutter to move to the extended position when the first roller rolls over the ramp. The second actuator also has a second roller. The second roller causes the second cutter to move to the extended position when the second roller rolls over the ramp. In an additional aspect, a diameter of the first roller is greater than a diameter of the second roller. In a further aspect, the first cutter is wider than the second cutter. In an additional aspect, an arc defined by the first cutter in the extended position as the shaft rotates is longer than an arc defined by the second cutter in the extended position as the shaft rotates. In a further aspect, a first receptacle is connected to the first cutter for receiving the portion of the interior wall of the tube cut by the first cutter, and a second receptacle is connected to the second cutter for receiving the sample cut by the second cutter. In an additional aspect, at least one spring is connected to the first cutter for biasing the first cutter against the interior wall of the tube when the first cutter is in the extended position, and at least one spring is connected to the second cutter for biasing the second cutter against the interior wall of the tube when the second cutter is in the extended position. In a further aspect, the first cutter is disposed opposite the second cutter. In an additional aspect, at least one spring is connected between the first cutter and the second cutter. The at least one spring biases the first and second cutters away from each other. In a further aspect, a motor is disposed in the cylindrical body and is operatively connected to the shaft for rotating the shaft. In another aspect, a tool for obtaining a sample from an interior wall of a tube has a cylindrical body having a central axis, an aperture in the cylindrical body, a shaft disposed in the cylindrical body along the central axis, an extension ramp connected to the cylindrical body, and a retraction ramp connected to the cylindrical body. A first cutter is operatively connected to the shaft for rotation therewith. The first cutter is movable radially between a retracted position where the first cutter is disposed inside the cylindrical body at a first distance from the central axis and an extended position where the first cutter extends at least in part through the aperture at a second distance from the central axis. The second distance is greater than the first distance. A first actuator is operatively connected to the first cutter for moving the first cutter between the retracted position and the extended position by interacting with the retraction ramp and the extension ramp respectively as the shaft rotates. A second cutter is operatively connected to the shaft for rotation therewith and is disposed at an angle to first cutter. The second cutter is movable radially between a retracted position where the second cutter is disposed inside the cylindrical body at a third distance from the central axis and an extended position where the second cutter extends at least in part through the aperture at a fourth distance from the central axis. The fourth distance is greater than the third distance. The fourth distance is greater than the second distance. A second actuator is operatively connected to the second cutter for moving the second cutter between the retracted position and the extended position by interacting with the retraction ramp and the extension ramp respectively as the shaft rotates. The second cutter is in the retracted position when the first cutter is in the extended position. The first cutter is in the retracted position when the second cutter is in the extended position. Rotating the shaft causes the first cutter to move to the extended position thereby cutting a portion of the interior wall of the tube and then causes the second cutter to move to the extended position thereby cutting the sample from the interior wall of the tube from a location in the tube revealed by cutting the portion of the interior wall of the tube. In an additional aspect, the first actuator includes a first actuation bar disposed generally parallel to the central axis. The first actuation bar has a first roller at a first end thereof, a second roller at a second end thereof, and at least one third roller between the first and second ends thereof. The second actuator includes a second actuation bar disposed generally parallel to the central axis. The second actuation bar has a fourth roller at a first end thereof, a fifth roller at a second end thereof, and at least one sixth roller between the first and second ends thereof. The extension ramp extends generally parallel to the central axis toward the first cutter and the second cutter, and defines an arc about the central axis. The retraction ramp extends generally parallel to the central axis toward the extension ramp, the first cutter and the second cutter, and defines an arc about the central axis. The first and second cutters are disposed between the extension ramp and the retraction ramp in a direction parallel to the central axis. A first holder is connected to the first cutter. The first holder has at least one slot defined therein at an angle to the central axis. The at least one slot of the first holder receives the at least one third roller therein. A second holder is connected to the second cutter. The second holder has at least one slot defined therein at an angle to the central axis. The at least one slot of the second holder receives the at least one sixth roller therein. When the first roller rolls over the extension ramp, the at least one third roller moves in the at least one slot of the first holder causing the first holder to move radially away from the central axis thereby causing the first cutter to move to the extended position. When the second roller rolls over the retraction ramp, the at least one third roller moves in the at least one slot of the first holder causing the first holder to move radially toward the central axis thereby causing the first cutter to move to the retracted position. When the fourth roller rolls over the extension ramp, the at least one sixth roller moves in the at least one slot of the second holder causing the second holder to move radially away from the central axis thereby causing the second cutter to move to the extended position. When the fifth roller rolls over the retraction ramp, the at least one sixth roller moves in the at least one slot of the second holder causing the second holder to move radially toward the central axis thereby causing the second cutter to move to the retracted position. In a further aspect, the extension ramp has a first ramp portion and a second ramp portion. The first ramp portion is longer than the second ramp portion. The first roller rolls over the first ramp portion of the extension ramp and the fourth roller rolls over the second ramp portion of the extension ramp. In an additional aspect, the retraction ramp has a first ramp portion and a second ramp portion. The first ramp portion is longer than the second ramp portion. The second roller rolls over the second ramp portion of the retraction ramp and the fifth roller rolls over the first ramp portion of the retraction ramp. In a further aspect, the first cutter is wider than the second cutter. In an additional aspect, an arc defined by the first cutter in the extended position as the shaft rotates is longer than an arc defined by the second cutter in the extended position as the shaft rotates. In a further aspect, a first receptacle is connected to the first cutter for receiving the portion of the interior wall of the tube cut by the first cutter, and a second receptacle connected to the second cutter for receiving the sample cut by the second cutter. In an additional aspect, at least one spring is connected to the first cutter for biasing the first cutter against the interior wall of the tube when the first cutter is in the extended position, and at least one second spring is connected to the second cutter for biasing the second cutter against the interior wall of the tube when the second cutter is in the extended position. In a further aspect, a motor is disposed in the cylindrical body and is operatively connected to the shaft for rotating the shaft. Embodiments of the present invention each have at least one of the above-mentioned objects and/or aspects, but do not necessarily have all of them. It should be understood that some aspects of the present invention that have resulted from attempting to attain the above-mentioned objects may not satisfy these objects and/or may satisfy other objects not specifically recited herein. Additional and/or alternative features, aspects, and advantages of embodiments of the present invention will become apparent from the following description, the accompanying drawings, and the appended claims. The circumferential sampling tool of the present invention will be described as being used for obtaining samples from pressure tubes of nuclear reactors to be analyzed for deuterium content. However it should be understood that the circumferential sampling tool could be used to collect other types of samples from other types of tubes or arcuate surfaces. Turning to FIGS. 2 and 3, an embodiment of a circumferential sampling tool 20 will be described. The tool 20 has a cylindrical body 22 having a central axis 24. The cylindrical body 22 has a plurality of bearing pads 26 for supporting the tool 20 when the tool 20 is disposed inside the pressure tube. An aperture 28 is defined in the cylindrical body 20. A cutter assembly 30, described in greater detail below, is disposed inside the cylindrical body 20 in longitudinal alignment with the aperture 28. The cutter assembly 30 is held by a carriage 32. The carriage 32 is connected via a coupler 34 to an output shaft 36 of an electric motor 38. The electric motor 38 is used to rotate the cutter assembly 30 as will be described in greater detail below. The electric motor 38 is preferably a DC motor, however other types of motors are contemplated. It is contemplated that the motor 38 could be coupled to the cutter assembly 30 differently. For example, the output shaft 36 of the motor 38 could be connected to a driveshaft which in turn is connected to the carriage 32. A purge tube 40 is connected to the cylindrical body 22. The purge tube 40 is used to dry the surface of a pressure tube where a sample is to be collected as described below. The circumferential sampling tool 20 is part of a circumferential sampling system, some of the features of which will be described briefly. The tool 20 is connected to a positioning system which permits accurate axial and angular positioning of the tool 20 in the pressure tube. A shielding sleeve is disposed over the tool 20 when the tool 20 is not pushed inside a pressure tube, thus closing the aperture 28. The tool 20, the positioning system, and the shielding sleeve are disposed on a support cart, which is preferably wheeled to facilitate the position of the cart. To obtain a sample from the interior wall of a pressure tube (including a rolled joint region), the cart is first rolled in position adjacent an opened end of the emptied tube. The opened end of the tube has an end fitting disposed thereon. The shielding sleeve is then connected to the end fitting. The positioning system is used to set the angular and axial position where the sample is to be collected inside the tube. As will be understood from the description of the cutter assembly 30 provided below, the cutter assembly 30 uses gravity to collect the sample, and therefore the sample is normally collected from the upper half of the tube (i.e. between the 9 o'clock and 3 o'clock positions). The tool 20 is then pushed inside the tube such that the cutter assembly 30 is past the location where the sample is to be collected. An air purge operation is then performed using the purge tube 40 to dry the location where the sample is to be collected. The tool 20 is then moved back inside the tube such that the cutter assembly 30 is aligned with the location where the sample is to be collected. The tool 20 is locked in this position and the bearing pads 26 are actuated to maintain the tool 20 in position by pushing against the interior wall of the tube. The motor 38 is then actuated, thus causing the cutter assembly 30 to rotate about the central axis 24. As it rotates, the cutter assembly cuts a portion of the interior wall of the tube in a circumferential direction thereof, thus obtaining the sample. Additional details regarding this step will be provided below when describing the cutter assembly 30. The tool 20 is then unlocked, the bearing pads 26 released, and the tool 20 retracted back inside the shielding sleeve. The sample contained in the cutter assembly 30 is then transferred to a flask contained in the cart. The above steps (starting with the setting of the angular and axial position where the sample is to be collected) can be repeated for obtaining other samples in other locations in the tube. Once all samples have been collected, the shielding sleeve is disconnected from the end fitting and the cart is rolled away from the pressure tube. Finally, the flask(s) containing the sample(s) is (are) retrieved. The above steps relate to one possible method of delivering the tool 20 inside a pressure tube to obtain samples. It should be understood that other methods of delivering the tool 20 are possible and contemplated. Turning now to FIGS. 3 to 6, the cutter assembly 30 will be described. The cutter assembly includes an oxide cutter 50 and a sample cutter 52 disposed opposite to each other. It is contemplated that the oxide cutter 50 and the sample cutter 52 could be disposed at other angles to each other. For example, it is contemplated that the oxide cutter 50 and the sample cutter 52 could be disposed perpendicularly to each other. The oxide cutter 50 and the sample cutter 52 are preferably made of carbide. The oxide cutter 50 is wider than the sample cutter 52 for reasons explained further below. The oxide cutter 50 is connected by a threaded fastener 54 to an oxide cutter cartridge 56. A chip clip 58 is connected to the oxide cutter cartridge 56. The chip clip 58 retains the portion of the tube being cut by the oxide cutter 50 inside a receptacle 60 formed between the oxide cutter 50, the oxide cutter cartridge 56, and the chip clip 58, as will be explained below. The oxide cutter cartridge 56 is connected by a bayonnet-type mount 62 to an oxide cartridge holder 64. Similarly, the sample cutter 52 is connected by a threaded fastener 66 to a sample cutter cartridge 68. A chip clip 70 is connected to the sample cutter cartridge 68. The chip clip 70 retains the sample being cut by the sample cutter 52 inside a receptacle 72 formed between the sample cutter 52, the sample cutter cartridge 68, and the chip clip 70, as will be explained below. The sample cutter cartridge 68 is connected by a bayonnet-type mount 74 to a sample cartridge holder 76. Two stacks of Belleville springs 78 are disposed between the oxide cartridge holder 64 and the sample cartridge holder 76, thus biasing the two cutters 50, 52 away from each other. A threaded fastener 80 is inserted in the sample cartridge holder 76 and abuts the oxide cartridge holder 64, thus retaining the springs 78 between the two holders 64 and 76. It is contemplated that other types of springs could be used instead of the Belleville springs 78. As will be described below, the cutters 50 and 52 are each movable (with the rest of the cutter assembly 30) between a retracted position where they are disposed inside the cylindrical body 22 and an extended position where they extend in part through the aperture 28 to cut the interior wall of the tube. The actuator for the oxide cutter 50 consists of two rollers 82 connected to either side of the sample cutter cartridge 68 (see FIG. 3) and of four springs 84. The rollers 82 are used to move the oxide cutter 50 to its extended position as will be described below. Two of the springs 84 are connected to the oxide cartridge holder 64 via two spring caps 86 and two of the springs 84 are connected to the sample cartridge holder 76 via two spring caps 86 (see FIG. 5). The springs 84 bias the oxide cutter 50 toward its retracted position. The actuator for the sample cutter 52 consists of two rollers 88 connected to either side of the oxide cutter cartridge 56 (see FIG. 3) and of the four springs 84. The rollers 88 are used to move the sample cutter 52 to its extended position as will be described below. The springs 84 bias the sample cutter 52 toward its retracted position. As can be seen in FIG. 3, the diameter of the rollers 82 is greater than the diameter of the rollers 88 for reasons discussed below. It is contemplated that the oxide and sample cutters 50, 52 could be actuated by other types of actuators. For example, it is contemplated that the rollers 82, 88 could be replaced by fixed cams. As best seen in FIG. 6, a ramp 90 is disposed inside the cylindrical body 22 along a circumferential portion thereof. As can be seen, the ramp 90 is disposed opposite the aperture 28. As discussed below, the roller 82, 88 roll over the ramp 90 to move the cutters 50, 52 to their extended positions. The method by which the cutter assembly 30 cuts the sample to be analyzed from the interior wall of the tube will now be described. The motor 38 turns the carriage 32 in the direction indicated by the arrow 92 in FIGS. 4 and 6, thus turning the cutter assembly 30 in the same direction. When the rollers 82 roll over the ramp 90, the cutter assembly 30 moves upwardly, thus moving the oxide cutter 50 to its extended position through the aperture 28. As the rollers 82 roll over the ramp 90, the oxide cutter 50 moves in an arc along a circumference of the interior wall of the tube and cuts an oxide layer from the interior wall of the tube. In a preferred embodiment, the oxide cutter 50 cuts slightly deeper than the oxide layer to ensure complete removal of oxide. The chip clip 58 causes the chip of oxide layer to curl inside the receptacle 60 as it is being cut. The Belleville springs 78 bias the oxide cutter 50 against the surface of the tube thus providing a cutting force, permitting the cutter to maintain contact with the surface should the surface be uneven and allowing the tool 20 to be used in a variety of pressure tube diameters. When the rollers 82 pass the ramp 90, the springs 84 bias the cutter assembly 30 back toward the inner wall of the cylindrical body 22, and therefore the oxide cutter 50 back to its retracted position. Once the oxide cutter 50 no longer contacts the interior wall of the tube, the chip of oxide layer falls inside the receptacle 60. The cutter assembly 30 then continues to rotate. When the rollers 88 roll over the ramp 90, the cutter assembly 30 moves upwardly, thus moving the sample cutter 52 to its extended position through the aperture 28. As the rollers 88 roll over the ramp 90, the sample cutter 52 moves in an arc along a circumference of the interior wall of the tube and cuts a sample from the interior wall of the tube from the location in the tube where the oxide layer was cut. The chip clip 70 causes the sample chip to curl inside the receptacle 72 as it is being cut. The Belleville springs 78 bias the sample cutter 52 against the surface of the tube thus providing a cutting force, permitting the cutter to maintain contact with the surface should the surface be uneven and allowing the tool 20 to be used in a variety of pressure tube diameters. When the rollers 88 pass the ramp 90, the springs 84 bias the cutter assembly 30 back toward the inner wall of the cylindrical body 22, and therefore the sample cutter 52 back to its retracted position. Once the sample cutter 52 no longer contacts the interior wall of the tube, the sample chip falls inside the receptacle 72. The sample cutter 52, when in the position shown in FIGS. 4 and 6, is disposed further from the central axis 24 than the oxide cutter 50, thus resulting in the sample cutter 52 cutting deeper than the oxide cutter 50. Therefore, as can be seen in FIG. 7, the depth Ds of the cut made by the sample cutter 52 is greater than the depth Do of the cut made by the oxide cutter 50. Also, as previously mentioned, the sample cutter 52 is narrower than the oxide cutter. Therefore, as can also be seen in FIG. 7, the width Ws of the cut made by the sample cutter 52 is smaller than the width Wo of the cut made by the oxide cutter 50. Since the rollers 88 have a smaller diameter than the rollers 82, the arc defined by the sample cutter 52 as it moves against the surface of the interior wall of the tube is shorter than the arc defined by the oxide cutter 50 as it moves against the surface of the interior wall of the tube. Therefore the sample chip is shorter than the oxide layer chip. The deeper, narrower, and shorter cut made by the sample cutter 52 ensures that the sample is free of oxide thus ensuring a reliable analysis of the deuterium concentration of the sample which can be used to determine the useful life of the pressure tube. Also, since the cutters 50, 52 move about the circumference of the interior wall of the tube, they are not affected by surface variations in the axial direction of the tube. Therefore, the tool 20 can be used to obtain samples in the rolled joint region of the pressure tube. Turning now to FIGS. 8 to 11, an alternative embodiment of the circumferential sampling tool 20 (circumferential sampling tool 120) will be described. For simplicity, features of the tool 120 which are similar to those of the tool 20 have been labelled with the same reference numerals and will not be described again in detail The tool 120 is provided with a cutter assembly 130. The cutter assembly 130 is rotated by a driveshaft 100 connected to the motor 38 (not shown in this embodiment). As can be seen in FIG. 8, the cutter assembly 130 has two oxide cutters 50A, 50B and two sample cutter 52A, 52B. The two oxide cutters 50A, 50B are disposed opposite to each other. Similarly the two sample cutters 52A, 52B are disposed opposite to each other. The sample cutters 52A, 52B are disposed perpendicularly to the oxide cutters 50A, 50B. Each of the cutters 50A, 50B, 52A, and 52B is connected to a corresponding cutter cartridge and a cartridge holder and has a corresponding chip clip, receptacle, and fasteners as in the cutter assembly 30. Therefore, for simplicity, these elements have been labelled with the same reference numerals as in the cutter assembly 30 with the addition of the corresponding suffix A or B, as the case may be, and will not be described again in detail. Each of the cutters 50A, 50B, 52A, and 52B is movable radially between a retracted position where it is disposed inside the cylindrical body 22 and an extended position where it extends in part through the aperture 28 to cut the interior wall of the tube. In a preferred embodiment, the distance between the oxide cutter 50A and the central axis 24 in its retracted and extended positions corresponds to the distance between the oxide cutter 50B and the central axis 24 in its retracted and extended positions, and the distance between the sample cutter 52A and the central axis 24 in its retracted and extended positions corresponds to the distance between the sample cutter 52B and the central axis 24 in its retracted and extended positions. In FIG. 8, the cutters 50B, 52A, and 52B are in their respective retracted position and the cutter 50A is in its extended position. Each of the cutters 50A, 50B, 52A, and 52B is provided with an actuator to move it between the two positions as the cutter assembly rotates. Except as otherwise indicated, the actuators for each of the cutters 50A, 50B, 52A, and 52B are the same and actuate the cutters 50A, 50B, 52A, and 52B in the same way. Therefore only the actuator of the oxide cutter 50A will be described in detail. As seen in FIG. 9, the actuator of the oxide cutter 50A includes an actuation bar 132A disposed generally parallel to the central axis 24. The actuation bar 132A has a roller 134A at a first end thereof, a roller 136A at a second end thereof, and two rollers 138A between the two ends thereof. As can be seen, the roller 134A is wider than the roller 136A, for reasons explained below. In the actuator for the sample cutters 52A and 52B, the relative roller width of the end rollers is the opposite (i.e. the roller corresponding to the roller 134A is narrower than the roller corresponding to the roller 136A), for reasons explained below. The two rollers 138A are received in two slots 140A formed in the lower portion of the oxide cartridge holder 64A. As can be seen, the two slots are disposed at an angle to the central axis 24. The actuation bar 132A is made in two parts 142A and 144A. The part 142A is received inside the part 144A and can move axially relative to the part 144A. Stacks of Belleville springs 146A bias the two parts 142A, 144A away from each other. An extension ramp 150 is disposed inside of and is connected to the upper portion of the cylindrical body 22. As can be seen in FIG. 10, the extension ramp 150 defines an arc about the central axis 24. The extension ramp 150 extends generally parallel to the central axis 24 toward the cutter assembly 130. As discussed below, the roller 134A rolls over the extension ramp 150 to move the oxide cutter 50A to its extended position. A retraction ramp 152 is disposed inside of and is connected to the lower portion of the cylindrical body 22. As can be seen in FIG. 12, the retraction ramp 152 defines an arc about the central axis 24. The retraction ramp 152 extends generally parallel to the central axis 24 toward the cutter assembly 130 and the extension ramp 150. As discussed below, the roller 136A rolls over the retraction ramp 152 to move the oxide cutter 50A to its extended position. As can be seen in FIGS. 10 and 11, the extension ramp 150 has long ramp portions 154 and a short ramp portion 156. The wide rollers 134A, 134B of the actuators of the oxide cutters 50A, 50B roll over the long ramp portions 154. The corresponding rollers of the actuators of the sample cutters 52A, 52B, which are narrow, roll over the short ramp portion 156. As can be seen in FIGS. 12 and 13, the retraction ramp 152 is longer than the extension ramp 150, and similarly has long ramp portions 155 and a short ramp portion 157. The narrow rollers 136A, 136B of the actuators of the oxide cutters 50A, 50B roll over the short ramp portion 157 of the retraction ramp 152. The corresponding rollers of the actuators of the sample cutters 52A, 52B, which are wide, roll over the long ramp portions 155 of the retraction ramp 152. As the motor 38 turns the cutter assembly 130 in the direction indicated by arrow 158 in FIG. 8, the roller 134A rolls over the long portions 154 of the extension ramp 150. This causes the actuation bar 132A to move axially towards the left in FIG. 9. As the actuation bar 132A moves left (as seen in FIG. 9), the rollers 138A push against the slot 140A, which, due to their angle relative to the central axis, cause the oxide cartridge holder 64A to move upwardly. Therefore, the oxide cutter 50A moves to its extended position through the aperture 28. As the roller 134A rolls over the ramp 150, the oxide cutter 50A moves in an arc along a circumference of the interior wall of the tube and cuts an oxide layer from the interior wall of the tube. In a preferred embodiment, the oxide cutter 50A cuts slightly deeper than the oxide layer to ensure complete removal of oxide. The chip clip 58A causes the chip of oxide layer to curl inside the receptacle 60A as it is being cut. The Belleville springs 146A bias the oxide cutter 50A against the surface of the tube thus providing a cutting force, permitting the cutter to maintain contact with the surface should the surface be uneven and allowing the tool 120 to be used in a variety of pressure tube diameters. As the motor 38 continues to rotate, the roller 136A rolls over the short portion 157 of the retraction ramp as the roller 134A rolls off the extension ramp 150A. This causes the actuation bar 132A to move axially towards the right in FIG. 9. As the actuation bar 132A moves right (as seen in FIG. 9), the rollers 138A push against the slot 140A, which, due to their angle relative to the central axis 24, cause the oxide cartridge holder 64A to move downwardly. Therefore, the oxide cutter 50A moves to its retracted position. Once the oxide cutter 50A no longer contacts the interior wall of the tube, the chip of oxide layer falls inside the receptacle 60A. As the motor 38 continues to rotate, the actuator of the sample cutter 52A moves the sample cutter 52A between its extended and retracted position in a similar manner. However, since the roller of this actuator rolls over the short portion 156 of the extension ramp 150, the arc defined by the sample cutter 52A as it moves against the surface of the interior wall of the tube is shorter than the arc defined by the oxide cutter 50A as it moves against the surface of the interior wall of the tube. Therefore the sample chip is shorter than the oxide layer chip. As in the cutter assembly 30, the sample cutter 52A also makes a narrower and deeper cut than the oxide cutter 50A. This is achieved by providing shims (not shown) between the sample cutter 52A and the sample cutter cartridge 68A. The deeper, narrower, and shorter cut made by the sample cutter 50A ensures that the sample is free of oxide thus ensuring a reliable analysis of the deuterium concentration of the sample which can be used to determine the useful life of the pressure tube. Once the sample has been cut by the sample cutter 52A and the sample cutter 52A has been returned to its retracted position, the motor 38 is stopped. The tool 120 is then repositioned in the pressure tube in order to obtain a second sample from a different location. Once the tool 120 is repositioned, the motor 38 is turned on so as to continue to rotate the cutter assembly 130 which causes the oxide cutter 50B to cut another oxide chip and the sample cutter 52B to cut another sample in the same manner as the one described above with respect to cutters 50A and 52A. Therefore, the tool 120 advantageously allows two samples to be cut before the tool 120 has to be retracted back inside the shielding sleeve to transfer the samples to flasks contained in the cart, whereas this step needs to be done after each sample is cut with the tool 20. Since the cutters 50A, 50B, 52A, and 52B move about the circumference of the interior wall of the tube, they are not affected by surface variations in the axial direction of the tube. Therefore, the tool 120 can be used to obtain samples in the rolled joint region of the pressure tube. Modifications and improvements to the above-described embodiments of the present invention may become apparent to those skilled in the art. The foregoing description is intended to be exemplary rather than limiting The scope of the present invention is therefore intended to be limited solely by the scope of the appended claims. |
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044951469 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT The present invention is a system and a method for loading nuclear fuel rods with spherical nuclear fuel. The overall system is shown in FIGS. 1 and 3. FIG. 1 shows the system viewed from the front. A fuel cladding rod 2 to be loaded with fuel is held vertically upright by a fuel rod support clamp 4. Because of the length of the fuel rods 2, the rods 2 may be set in pit 1 in the building floor. The support clamp 4 is fixed to a vibrator 6 driven by vibrator motor 8. The vibrator 6 rests on a frame 10. The frame 10 is vertically adjustable to give the vibrator 6 a vertical travel of several feet. This allows the loading system to accomodate fuel cladding rods 2 of different lengths. The support clamp 4 is a cage structure of six vertical rods 12, three of which are held in a spaced parallel array by semi-circular pieces 14 (see FIG. 2). The semi-circular pieces 14 are hinged at one end 16 and clamped 18 at the other. Only one of the bottom semicircular pieces 14 of the support clamp 4 is fixed to the vibrator 6. This allows the half of the support clamp 4 whose bottom semi-circular piece 14 is not fixed to the vibrator 6 to swing open about the hinge 16 disclosing the interior of the clamp. This arrangement allows easy horizontal access to the interior of the fuel rod support clamp 4 in order to insert empty fuel cladding tubes 2 and remove loaded tube 2. The open upper end of the fuel cladding tube 2 is attached to an adaptor 18 with an airtight connection. The adaptor 18 is mounted to the glove box 22 via a bellows arrangement so that the fuel tube 2 is flexibly mounted to the glove box 22 allowing the fuel tube 2 to vibrate in response to the vibrator 6 while the tube is being locked. The adaptor 18 is connected to the glove box 22 with a vacuum valve 24 so that the adaptor 18 and fuel tube 2 combination may be isolated from the glove box 22 forming an airtight combination. The glove box 22 is an enclosure capable of being made airtight which receives the nuclear fuel through the entrance vacuum valve 26. The glove box includes windows 28 and hinged glove box covers 30. Opening the glove box covers 30 reveals gloves (not shown) mounted to the glove box 22 which allows the operator to accomplish manipulation within the glove box 22 while still retaining the inert atmosphere within the glove box 22. On the upper side and connected to the glove box 22 is the rod loading assembly cover 32. The rod loading assembly cover 32 is of sufficient length to allow the rod loading assembly 34 to rise high enough so that it is free of the fuel cladding tube 2. FIG. 3 shows a side view of the glove box and fuel cladding tube assembly. After entering the glove box through the vacuum valve 26, the fuel proceeds to the weighing station. Referring to FIGS. 4, 5, 6, and 7 shows the passage of the spherical nuclear fuel from the entrance vacuum valve 26 to the loading hoppers 60 and 62 of the weighing stations. The nuclear fuel spheres enter the glove box 22 through the entrance vacuum valve 26 in containers 36 large enough to hold sufficient fuel for about six fuel tubes 2. The fuel containers 36 indicated by the dotted lines, move along the rollers 40 of the transport conveyer 38, which may be powered or non-powered. After coming to rest on the conveyor 38, the fuel is lifted vertically upward by the overhead transport system 42. The over head transport system 42 is capable of lifting the fuel container 36 from the transport conveyor 38 and moving it from right to left and back and forth within the glove box 22. The overhead transport conveyor 38 includes a rotating drum 41 around which is wrapped a cord 44 for raising and lowering the spheres. The containers 36 are moved one at a time from the transport conveyor 38 to the loading hoppers 60 and 62 of the weighing scales. In the preferred embodiment, three sizes of spheres are used, which are referred to herein as fines, mediums and large. There are three weighing stations, one corresponding to each of the sphere sizes. However, only two of the weighing stations, the fines 64 and mediums 66 are shown for clarity. The mediums weighing station 66 is shown in FIG. 4 by the dotted figure in the load position for receiving fuel. The fuel container 36 is attached to the transport lid 43 and moved by the overhead transport system 42 to each of the weighing stations where the fuel spheres are deposited into the hoppers 60 and 62 of the scales. The mediums 66 and large 65 weighing stations are mounted on one platform and move from side to side by the drive motor 68. In addition, the weighing stations move up and down by the drive mechanism 72, the glove box 22 providing a recess 74 for the support shaft 76 when the station is lowered. The fines weighing station 64 moves front to back driven by the drive motor 78 within the glove box 22 as well as side to side motion driven by motor 70. FIG. 7 shows a schematic view of the weighing station system viewed from above. The relative motion of the weighing stations, fines 64, mediums 66 and large 65 is indicated by arrows. The loading of the weighing station hoppers occurs when the stations are located at the position L of FIG. 7. As noted above, the mediums 66 and large 65 weighing stations are mounted on one platform and are moved to position L in FIG. 7 for loading. In addition, the weighing stations loading hoppers 61 and 62 are lowered to accommodate the fuel containers 36 which are moved to the weighing stations by the overhead fuel transport 42. The fines weighing system 64 is mounted independently of weighing stations 65 and 66 and moves toward the back of the glove box 22, then to the right and down for loading. The fuel spheres containers 36 are picked up by the overhead transport system 42 and positioned on top of the weighing station hoppers. The spheres are released into the loading hoppers 60, 61 and 62. Spheres of each size are dropped into the weighing scales hoppers 80 and 82 in incremental amounts by the stepper motors 84 and 86. When predetermined amounts of fuel spheres are received by the scale hoppers 80 and 82 as indicated by the weighing means 79 and 81 the flow ceases (recall only two of the three weighing scales are shown in the Figures). These three predetermined amounts of the three sizes of fuel spheres are sufficient to fill one fuel rod 2 when they mixed together within the fuel rod. These fuel spheres are then transferred to the hoppers 92, 94, and 96 of the feeding probe 34. The feeding probe 34 is described in more detail in copending application Ser. No. 327,816 entitled "Spherical Nuclear Fuel Loading Probe" filed by the same inventor and assigned to same assignee as the present application and incorporated herein by reference. The probe is a means for feeding the three different sizes of spheres into the fuel rod 2 in a controlled manner so that the correct uniform density is achieved in the rod 2. Referring to FIGS. 8, 9 and 10, the probe 34 includes three funnels 92, 94 and 96 into which each of the three quantities of fuel is discharged from the weighing scale hoppers 80, 81 and 82. In FIG. 8, only two 92 and 94 of the three funnels are shown for clarity. The funnels are spaced about 60.degree. apart and are all identical except for the ability to accommodate different sized spherical fuel. The three funnels 92, 94 and 96 are connected to the probe hopper 98 via three solenoid valves 100. There is one solenoid valve 100 for each funnel. The probe hopper 98 is divided into three sections 102, 104 and 106 (see FIG. 7). The fuel spheres, after being released by the solenoid valves 100, pass through a regulator gate 114 shown in FIGS. 8, 9 and 11. The gate 114 is releasably attached by conventional ball plunger means 115 to the hopper 98 so as to restrict the passageway 116 connecting the funnels 92 to the sections 102, 104 and 106 of the probe hopper 98. The gate 114 includes an opening of height h and width w. These dimensions are selected according to the size of nuclear fuel spheres and the desired rate of flow into the probe hopper 98. The rate of flow of each of the fuel spheres is determined so that upon emergence from the probe 34 within the fuel rod 2, the maximum randomness of the three different size spheres is achieved. The probe hopper 98 is connected to tubing 108, 110 and 112. Each of these tubes corresponds to one of the sections of the probe hopper 98 which, in turn, corresponds to one of the funnels 92, 94 and 96. In the particular embodiment shown in FIG. 8, two of the tubes 108 and 110 are of the same circular cross-section. These tubes are used for the two smallest diameter fuel. The largest fuel sphere is carried by the tube 112 of larger cross section. The outer surface of lower end 115 of each of the tubes 108, 110 and 112 may be extended into scoops as shown in FIG. 8. The scoop shaped extensions terminate toward the axis through the center of the three tubed arrangement. These extensions help in the mixing of the fuel spheres to provide a random distribution of packing of the fuel tube. To further enhance the randomness of distribution of the three different sized spheres and improve the uniformity of packing of the fuel rod 2, a cone shaped piece 118 may be fixed to the lower end of the fuel tubes 108, 110 and 112 by two cylindrical rod members 130 as shown in FIG. 8. The cone 118 is fixed to the rods 130 by conventional means. The method for using the above described system is described below. Some of the operations described below are carried by the operator through the glove openings 28. Other operations are carried out automatically e.g., the raising and lowering of the fuel probe 34. The operator places an empty fuel rod 2 in the fuel rod support clamp 4 fixed to the vibrator 6. The upper end of the rod 2 is connected to the adaptor 18 with a flexible coupling to the glove box 22. The feeding probe 34, which is attached to a raising and lowering means 33 by a cable 35 attached to bracket 31 is lowered into the fuel rod 2 so that the tubing 108, 110 and 112 reaches to the bottom of the fuel rod 2. The operator then transfers the fuel sphere container 36 from the transport conveyer 40 to the loading position L (See FIG. 7) by the overhead transport system 42. At position L, the fuel spheres are transferred to the hoppers 60, 61 and 62 of the weighing scales. Predetermined amounts of fuel are released to the weighing station hoppers 80 and 82 by the stepper motors 84 and 86. The weighing stations are then moved into the dump position relative to the feed probe 34 so that the fuel spheres may be unloaded into the feeding probe hoppers. FIG. 4 shows the medium weighing station 66 in the loading position as indicated by the dotted outline. After loading the fuel into hopper 62, the weighing station is raised and moved into position as indicated by the solid figure to unload the fuel spheres into the fuel probe. FIGS. 5 and 6 show the fines weighing station 64 in the load position and the dump position, respectively. The weighing stations are moved out of the way of the fuel feeding probe 34 and the solenoid valves 100 are opened. The fuel spheres descend through the valve 100, the regulator gate 114 and the tubing 108, 110 and 112. As the fuel reaches the bottom of the fuel rod 2, the probe 34 is raised at a rate so that the bottom of the fuel probe 34 remains just above the ascending fuel column. That is, the spheres are deposited on top of the fuel column such that the end of the probe remains between about 1 and 5 in above the ascending fuel column. Copper tubes 132 guide the feeding probe 34 up and down. The copper tubes 132 in combination with wires 134 provide the electrical contact to operate the solenoids 100. The vibrator 6 is in operation while the fuel rod 2 is being loaded to assist in the packing of the fuel rod 2. After the loading is completed and feeding probe 90 is clear of the fuel rod 2, the rod is removed from the support clamp 4. A new fuel rod is placed in the clamp 4 and process is started again. |
049960172 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The neutron generating system and the associated improved neutron generator tube of the present invention can perhaps be best explained and understood by reference to the accompanying drawing. FIG. 1 illustrates a side view of a preferred neutron generator tube involving a cylindrically symmetric housing or sealed envelope generally designated by the numeral 10. The overall housing 10 is made up of an ion source enclosure 12 on the left, a frustro-conical high voltage insulator 14, a secondary electron emission suppressor electrical contact 16, a target insulator 18, and a target thermal conductor enclosure 20. Through the transparent high voltage insulator the polished exterior surface of the secondary electron emission suppressor 22 can be seen. On the right and extending out of the housing 10 is the target thermal conductor 24. On the left and also extending out of the housing 10 are a pair of anode electrical terminals 26 and 28, a gas pumping tube 30, and an ion source gas replenishing terminal 32. Centrally located in the end of the housing 10 and as illustrated in FIG. 2 is a cylindrical recess or hole 34 for accepting a removable magnet. FIG. 3 illustrates a partial cut-away view of a neutron generator tube, again generally designated by the numeral 10, very similar to the tube of FIG. 1 except a permanent magnet 36 and magnetic pole piece 38 are positioned within the recess 34. FIG. 3 has been rotated 180.degree. about the axis of the tube to illustrate the symmetry of the tube. The housing 10 can again be seen to be made up of an ion source enclosure 12, a high voltage insulator 14, a secondary electron emission suppressor electrical contact 16, a target insulator 18, and a target thermal conductor enclosure 20. As illustrated the ion source enclosure is made up of a header 40 and a hollow cylindrical tubular piece 42 that attaches to the larger end of the frustro-conical glass high voltage insulator 14. A pair of high voltage insulated anode electrical terminals 26 and 28 pass through the header 40 into the annular cavity 44 defined between the tubular piece 42 and the cylindrical recess wall 46 with end cap 48 adapted to accept the magnet 36 and magnetic pole piece 38 (see FIG. 4). The electrical terminals 26 and 28 then attach to and support a ring anode 50 adjacent to the inner side of the recess end cap 48 and axially aligned within the ion source enclosure 12. In this manner the removable magnet 36 can be advantageously positioned extremely close to the ion source chamber 52. To the right of the ring anode 50 and ion source chamber 52 is a three layered novel ion screen ensemble or shield 54 containing an axially positioned gridded aperture 56. This ion screen ensemble, see FIG. 5, is made up of an outer highly polished thin-walled cover 58 visible through the transparent high voltage insulator 14. On the inner or anode side of the ensemble 54 is a magnetic metal disc 60 that serves several purposes. Disc 60 lends structural support as well as tends to confine the magnetic field to the ion source chamber 52. Disc 60 in combination with the outer cover 58 serves to retain a thin metal grid 62 sandwiched between these pieces and covering the centrally located aperture 56 passing through both the disc 60 and cover 58. The multilayered screen structure is positioned very close to the ring anode such as to optimize the efficiency of extracting ions from the ion source chamber 52. The presence of the grid 62 across the aperture 56 tends to broaden the ion beam resulting in a wider beam of lower power per unit area. At the other end of the neutron tube is the target thermal conductor 24 in the form of a metallic cylindrical rod penetrating the housing end cap 64 and being held axially suspended within the smaller diameter end of the tube (see FIG. 7). The inner end surface of the conductor 24 is highly polished and contains a thin film of hydride forming metal (preferably titanium) deposited directly on the conductor 24 to serve as the cathode target 66. Displaced to the left of the target 66 is an axially aligned metallic secondary electron emission suppressor 68 containing an axially aligned aperture 70 (see FIG. 6). This suppressor 68 is electrically connected to the exterior suppressor enclosure 16 such that during operation of the neutron generator tube the suppressor 68 can be maintained at an electrical potential a few thousand volts below the cathode target 66 potential such as to attract any secondary electron emission originating from the face of the target as is generally known and practiced in the art. In selecting materials to build the neutron tube, the ion source enclosure 12 and components surrounding the anode ring should be constructed out of ferromagnetic materials such as to confine the magnetic field to the ion source chamber 52. Preferably magnetic stainless steel 410 or 17-4 is employed for this purpose. The metal to glass insulation junctions are preferably of Kovar to 7052. Thus, cylindrical piece 42 of the ion source enclosure 12, the secondary electron emission suppressor enclosure 16 and the target thermal conductor enclosure 20 are preferably made of Kovar. The stainless steel to Kovar junctions are preferably GTAW welds performed during assembly of the tube in the absence of the removable magnet. All electrical leads and the vacuum tube ports are hermetically sealed before evacuating the tube. An evacuation or bake out procedure is employed prior to filling the tube with an ionizable gas. This procedure conventionally involves pulling a vacuum on the gas pumping tube 30 while the entire neutron generator tube is being sustained for an extended period of time at an elevated temperature. After the bake out procedure the gas pumping tube 30 is employed to fill the interior of the tube with an ionizable gas such as deuterium or tritium. Preferably a 50/50 mixture of deuterium and tritium is employed. The gaseous mixture is then deposited on the gas replenishing device of FIG. 8 located in the annular cavity 44 in a manner well known in the art. As illustrated in FIG. 8, a heating element 72 surrounded by a ceramic insulator 74 and concentric exterior cylinder of titanium 76 is electrically attached at one end to the ion source gas replenishing terminal 32 after it passes through the header 40 and at the other end is grounded to the recess wall 46. By passing electrical current through the heating element 72, the temperature of the ceramic insulator 74 and titanium sleeve 76 are controlled such as to thermally react the titanium with the gaseous deuterium and tritium producing the hydrides (i.e., the titanium thermally absorbs the hydrogen gas). Since this reaction is thermally reversible the gas pressure can be regulated during the lifetime of the neutron generator tube by merely controlling the temperature of the titanium sleeve 76 containing the deposited hydrides. Having once achieved the desired deposition of the hydrides and proper pressure within the interior of the tube, the neutron generator tube is hermetically sealed by crimp welding the gas pumping tube 30. The neutron generator tube is now ready to be used. A permanent magnet and magnetic pole piece can be inserted into recessed hole 34 with confidence that the magnetic properties have not been deleteriously affected by the high temperatures associated with fabrication of the tube and the degassification/gassification processes. The external portion of the housing 10 is equipped with 0-ring grooves 78 to allow for sealing the tube within a downhole logging tool. In this sense, the overall device is extremely compact and compatible with even the smallest of borehole diameters yet excellent neutron flux densities can be achieved. In order to operate the tube, a high voltage source and appropriate control circuitry are attached to the anode electrical terminals 26 and 28 and cathode target conductor 24. As is known in the art, a high voltage of the order of 100 K volts or even greater is applied across the ring anode and target cathode, whereupon free electrons originating at the cathode are accelerated into the ion chamber, and in response to the presence of the magnetic field induced spiral path, collide and ionize the tritium and deuterium present in the ion chamber. The resulting tritium and deuterium nuclei (positive ions) are then extracted (by virtue of the electrical potential) from the ion chamber and are accelerated towards the cathode. Upon collision with the hydrogen-containing target (i.e., deuterium and tritium occluded titanium target) the desired neutron releasing nuclear reaction takes place; i.e., the d,d reaction emits neutrons of approximately 2.5 MeV and the d,t reaction emits neutrons of approximately 14 MeV. Conventionally for well logging purposes, the neutron emission is performed on a pulsed basis with approximately a 5 to 10 percent duty cycle. However, the neutron source of the present invention has also been successfully used in a sustained continuous neutron emission mode of operation. The use of a 50 percent deuterium and 50 percent tritium mixture as the ionizable gas is preferred in that this mixture is (as known in the art) a self-replenishing hydride source; however, other mixtures of the hydrogen isotopes are acceptable alternatives and should be considered equivalent for purposes of this invention. The target replenishing procedure in the present invention can be periodically performed by removing the magnet and placing the entire neutron generator tube in an oven at the appropriate temperature to form the deuterium and tritium hydrides of titanium. An electrical current is then used to heat the element 76 attached to the ion source gas replenishing terminal 32 (see FIG. 8) to a temperature characteristic of reversing the titanium hydride formation reaction. In this manner, net migration of occluded hydrides of deuterium and tritium from the titanium sleeve 76 to the titanium target 70 and reestablishing the desired ionizable gas pressure within the tube is accomplished, again without deleteriously affecting the permanent magnet. Thus the lifetime of the neutron generator tube can be extended to several hundred hours. An improved neutron tube, as illustrated in the drawing, having an external housing diameter of the order of about 27 mm and nominally designed to operate at borehole temperatures of 200.degree. C. with neutron flux of 10.sup.8 neutrons per second or greater, has been successfully tested at temperatures as high as 190.degree. C. with negligible thermal effects. In contrast to the prior art devices, no thermally induced degradation of the exothermal hydrogen absorber target was observed. Several critical features of the present invention are felt to contribute this success. The fact that the permanent magnet can be withdrawn from the neutron generator tube during fabrication, degassification, gassification and gas replenishing, allows for optimum preservation of permanent magnetic properties during periods when the tube is subjected to ultra-high temperatures. It further allows the use of highly temperature-sensitive magnets which otherwise would be inoperative. Thus, the concept of providing for a removable magnet of the present invention can be advantageously employed with generally any previously known permanent magnetic material as well as the preferred samarium/cobalt magnet. And, because of the optimization of the magnetic field strength, the overall physical dimensions of the tube can be minimized without sacrificing the efficiency of the ion source and the desired narrow pulse width of the neutron emission. This overall compactness of the tube is viewed as being critical and advantageous in the sense of making the tube capable of being used in even the smallest oil and gas well borehole. Good thermal contact of the target surface with the outside wall is considered necessary for the successful operation of the neutron generator tube. Thus the presence of the thermally conductive target support preferably having a cross-section of substantially the same size as the target wherein one end of the support contains the target and is within the tube and the other end extends outside the tube will advantageously assist in removing thermal energy from the target. Thus the target support can be fabricated from any thermally conductive inert metal. Preferably a low oxygen containing copper (OFHC) rod is employed. The target end is highly polished and a film of hydrogen absorbtive metal is deposited to serve as the target. Generally any of the well known metallic target materials can be employed for this purpose; however, a titanium film of about 5 to 10 microns deposited on the OFHC copper is preferred. Because of the above advantages, a temperature gradient between the target and the outside of the tube of about 30.degree. C. is anticipated at downhole operating conditions. The presence of the ion pervious grid in the ion screen ensemble near the anode ring of the Penning ion source is considered critical in that it affords a wide ion beam of reduced power per unit area impinging on the target. Consequently the occurrence of localized hot spots on the target is alleviated and the tendency toward thermal degradation of the target is diminished. The ion grid can be made of any inert conductive material structurally capable of spanning the axially oriented aperture. Preferably the screen itself is made of etched tungsten, and as previously mentioned, the screen preferably is sandwiched between a magnetic disc on the inside (ring anode side) and a thin highly polished metallic disc on the target side. It has been discovered that an etched tungsten film of about 0.002 to 0.005 inch thick with the openings representing about 90 percent of the surface area, covering about a 1/8th inch diameter aperture, and positioned within about 0.030 inch of the anode ring of a neutron generator tube as illustrated in the drawing will result in an ion extraction efficiency of better than 40 percent of the ion formed. It has further been demonstrated that ignition or light-off of the ion source will occur at as low as 300 volts potential drop between the ring anode and target cathode when the ion screen is grounded. It is felt that the electric field penetration of the screen and thus the time of ignition of the ion source can be controlled by providing the ion screen with a means to externally vary the relative potential of the ion screen. In this manner the pulsing neutron generator could be more accurately controlled than previously known methods involving timed ultrahigh voltage circuitry. Having thus described the preferred embodiments of the invention with a certain degree of particularity, it is manifest that many changes can be made in the details of the construction and the arrangement of components without departing from the spirit and scope of this disclosure. For example, the overall design and shape of the neutron generator housing does not have to be cylindrically symmetric about the longitudinal axis of the tube and the specific and relative sizes of the components thereof can be readily varied. Therefore, it is to be understood that the invention is not limited to the embodiments set forth herein for purposes of exemplification, but is to be limited only by the scope of the attached claims including a full range of equivalents to which each element thereof is entitled. |
052020843 | claims | 1. A nuclear reactor comprising: a vessel containing water; a first core disposed in said vessel and containing a plurality of first fuel bundles configured in a first two-dimensional array for boiling said water to form a steam and water mixture; a plurality of steam separators disposed above said first core for receiving said steam and water mixture and being effective for separating said water from said steam; and a second core disposed in said vessel and containing a plurality of second fuel bundles configured in a second two-dimensional array disposed above said steam separators for receiving and heating said separated steam to form superheated steam. an inlet disposed in flow communication with said respective first fuel bundle for receiving said steam and water mixture therefrom; a steam outlet disposed in flow communication with a respective one of said second fuel bundles for channeling thereto said separated steam; and a liquid outlet disposed in flow communication with said bypass inlets for channeling said separated water into said bypass channel. said bypass channels include a plurality of bypass outlets disposed in a lower support plate supporting said first core for channeling said separated water into a lower plenum disposed below said first core; and said crud separating means include a removal conduit disposed in flow communication with said lower plenum for removing a portion of said water therein including said separated water, a filter for removing said crud from said water, and a return conduit for returning said filtered water back to said lower plenum. a first portion disposed longitudinally coextensively with said first fuel bundles and containing a nuclear poison; a second portion disposed longitudinally coextensively with said steam separators and being inert; a third portion disposed longitudinally coextensively with said second fuel bundles and containing a nuclear poison; and said control rods being movable longitudinally for controlling reactivity of said first and second cores. 2. A nuclear reactor according to claim 1 wherein each of said first fuel bundles is disposed in a flow baffle, adjacent ones of said baffles being spaced from each other to define a bypass channel for receiving and channeling said separated water away from said steam separators. 3. A nuclear reactor according to claim 2 wherein each of said baffles extends upwardly above said first fuel bundle to define a skirt spaced around a respective one of said steam separators, said skirt including a plurality of circumferentially spaced bypass inlets for channeling said separated water from said steam separator into said bypass channel. 4. A nuclear reactor according to claim 3 wherein each of said steam separators includes: 5. A nuclear reactor according to claim 4 wherein said skirts extend upwardly into said second core to a predetermined height above said steam separators for maintaining a predetermined level of said water in said bypass channels to provide a differential head for promoting separation of said steam from said steam and water mixture in said steam separators. 6. A nuclear reactor according to claim 4 further including means for separating crud from said separated water flowing in said bypass channels. 7. A nuclear reactor according to claim 6 wherein: 8. A nuclear reactor according to claim 4 further including a plurality of control rods disposed in respective bypass channels and each including: 9. A nuclear reactor according to claim 8 further including a plurality of control rod drives joined to said control rods, respectively, for selectively moving said control rods for withdrawing between said first and second cores either said first or third portions thereof. 10. A nuclear reactor according to claim 8 wherein said control rod drives are disposed below said first core and are effective for moving downwardly said control rods for withdrawing said third portion to between said first and second cores, and said first portion to below said first core. 11. A nuclear reactor according to claim 10 wherein said first and second fuel bundles are substantially identical in configuration and aligned longitudinally. 12. A nuclear reactor according to claim 11 further including means for separating crud from said separated water flowing in said bypass channels. 13. A nuclear reactor according to claim 4 wherein said first and second fuel bundles are substantially identical in configuration and aligned longitudinally. |
claims | 1. A method to prevent stress corrosion cracking of a storage canister by applying a compressive stress to a range where a tensile residual stress is generated on a metallic cylindrical body by welding a cover to a top of the cylindrical body,the method comprising:applying a first compressive stress beforehand to the range of the cylindrical body where the tensile residual stress is expected to be generated by the welding of the cover;canceling the tensile residual stress generated by the welding of the cover, with a compressive residual stress generated in the range; andthen applying a second compressive stress so as to generate a compressive residual stress over the range. 2. The method to prevent stress corrosion cracking of a storage canister according to claim 1, wherein the range of the cylindrical body that receives the first compressive stress is an axial range extending inward from an upper end of the cylindrical body in an axial direction, the axial range L satisfying a relational expression below:L≧2.5√{square root over (rt)}(r: an external radius of the cylindrical body, t: a thickness of the cylindrical body). 3. The method to prevent stress corrosion cracking of a storage canister according to one of claims 1 and 2, wherein the first compressive stress is applied by one of zirconia shot peening and burnishing. 4. A storage canister comprising a metallic cylindrical body with a cover welded to a top of the cylindrical body, the storage canister being installed in a cask while containing nuclear fuel in a sealing state,wherein a first compressive stress is applied beforehand to a range of the cylindrical body where a tensile residual stress is expected to be generated by the welding of the cover, the tensile residual stress generated by the welding of the cover is canceled with a compressive residual stress generated in the range, and then a second compressive stress is applied so as to generate a compressive residual stress over the range. 5. The storage canister according to claim 4, wherein the second compressive stress is applied to an upper opening between the cask and the cylindrical body, allowing generation of the compressive residual stress over the range. 6. The storage canister according to claim 5, wherein the cover includes an upper cover welded to an upper end of the cylindrical body and a lower cover welded to the cylindrical body inside the upper cover, and the lower cover is welded at a position in an axial range from the upper end of the cylindrical body to an L minimum value indicated by a right side of a relational expression below:L≧2.5√{square root over (rt)}(r: an external radius of the cylindrical body, t: a thickness of the cylindrical body). |
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043691617 | abstract | A rack and pinion mechanism for vertical up-and-down movement of neutron-absorbing control elements to control the power and to shut down a nuclear reactor in case of emergency. The upper portions of the elements constitute the rack, engaging a rotatably driven pinion perpendicular thereto and having teeth on a lateral face engaging corresponding teeth of a clutch pinion. The rack and pinion remain on continuous enagement, while the clutch pinion selectively engages and disenages the pinion, the translatory axial movement of the clutch pinion being controlled by a vertically moving actuating member controllable from the upper part of the apparatus. |
claims | 1. A silicotitanate molded body comprising:crystalline silicotitanate particles that have a particle size distribution in which 90% or more, on volume basis, of the particles have a particle size within a range of 1 μm or more and 10 μm or less and that are represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; andan oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium. 2. The silicotitanate molded body according to claim 1, further comprising niobium. 3. The silicotitanate molded body according to claim 2, wherein the silicotitanate molded body has a compressive strength at failure of 5.0 N or more. 4. The silicotitanate molded body according to claim 1, wherein a content of the oxide of one or more elements selected from the group of aluminum, zirconium, iron, and cerium is 20 wt % or less. 5. The silicotitanate molded body according to claim 1, wherein the molded body has a cylindrical shape having an average diameter within a range of 300 μm or more and 3,000 μm or less. 6. An adsorbent for cesium and/or strontium, comprising the silicotitanate molded body according to claim 1. 7. A decontamination method of a radioactive waste solution, comprising bringing an adsorbent for cesium and/or strontium comprising the silicotitanate molded body according to claim 1 into contact with a waste solution containing radioactive cesium and/or radioactive strontium. 8. The decontamination method of a radioactive waste solution according to claim 7, comprising bringing the radioactive waste solution into contact with the adsorbent in a column flow mode at a linear velocity LV of 2 m/h or more and 40 m/h or less and a space velocity SV of 10 h−1 or more and 300 h−1 or less. 9. A production method of the silicotitanate molded body according to claim 1, comprising:extruding a mixture containing crystalline silicotitanate that has a particle size distribution in which 90% or more, on volume basis, of particles have a particle size within a range of 1 μm or more and 10 μm or less and that is represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; and an oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium to form a molded body; and subsequentlydrying the molded body. 10. An adsorbent for cesium and/or strontium, comprising the silicotitanate molded body according to claim 2. 11. An adsorbent for cesium and/or strontium, comprising the silicotitanate molded body according to claim 3. 12. An adsorbent for cesium and/or strontium, comprising the silicotitanate molded body according to claim 4. 13. An adsorbent for cesium and/or strontium, comprising the silicotitanate molded body according to claim 5. 14. A decontamination method of a radioactive waste solution, comprising bringing an adsorbent for cesium and/or strontium comprising the silicotitanate molded body according to claim 2 into contact with a waste solution containing radioactive cesium and/or radioactive strontium. 15. A decontamination method of a radioactive waste solution, comprising bringing an adsorbent for cesium and/or strontium comprising the silicotitanate molded body according to claim 3 into contact with a waste solution containing radioactive cesium and/or radioactive strontium. 16. A decontamination method of a radioactive waste solution, comprising bringing an adsorbent for cesium and/or strontium comprising the silicotitanate molded body according to claim 4 into contact with a waste solution containing radioactive cesium and/or radioactive strontium. 17. A decontamination method of a radioactive waste solution, comprising bringing an adsorbent for cesium and/or strontium comprising the silicotitanate molded body according to claim 5 into contact with a waste solution containing radioactive cesium and/or radioactive strontium. 18. A production method of the silicotitanate molded body according to claim 2, comprising:extruding a mixture containing crystalline silicotitanate that has a particle size distribution in which 90% or more, on volume basis, of particles have a particle size within a range of 1 μm or more and 10 μm or less and that is represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; and an oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium to form a molded body; and subsequentlydrying the molded body. 19. A production method of the silicotitanate molded body according to claim 3, comprising:extruding a mixture containing crystalline silicotitanate that has a particle size distribution in which 90% or more, on volume basis, of particles have a particle size within a range of 1 μm or more and 10 μm or less and that is represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; and an oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium to form a molded body; and subsequentlydrying the molded body. 20. A production method of the silicotitanate molded body according to claim 4, comprising:extruding a mixture containing crystalline silicotitanate that has a particle size distribution in which 90% or more, on volume basis, of particles have a particle size within a range of 1 μm or more and 10 μm or less and that is represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; and an oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium to form a molded body; and subsequentlydrying the molded body. 21. A production method of the silicotitanate molded body according to claim 5, comprising:extruding a mixture containing crystalline silicotitanate that has a particle size distribution in which 90% or more, on volume basis, of particles have a particle size within a range of 1 μm or more and 10 μm or less and that is represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; and an oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium to form a molded body; and subsequentlydrying the molded body. |
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claims | 1. A radiation imaging apparatus comprising:a radiation source for performing multiple radiation emissions to a subject;a filter having at least one filter area for changing energy distribution of radiation in at least one of the multiple radiation emissions, the filter periodically shifting between an inserted state in which the at least one filter area is inserted in a path of radiation and a retracted state in which the at least one filter area is retracted from the path;a cycle determining section for determining a shift cycle of the filter for shifting the at least one filter area between the inserted state and the retracted state before a start of the first radiation emission based on subject information;a driver for driving the filter to shift at the determined shift cycle, the driver driving the filter to start shifting before a start of the first radiation emission and to shift until an end of the last radiation emission after the first radiation emission; andan emission controller for outputting a signal for starting a radiation emission to the radiation source to control emission timing of radiation,wherein the subject information includes cardiac cycle information and respiration information of the subject, and the radiation imaging apparatus further includes a signal detector for detecting a cardiac signal indicating a state of the cardiac cycle information and a respiratory signal indicating a state of the respiration information, andwherein the cycle determining section determines the shift cycle so as to be in synchronization with the cardiac cycle information, and the emission controller synchronizes the emission timing with a phase of the respiration based on the respiratory signal. 2. The radiation imaging apparatus of claim 1, further including a filter phase detector for detecting a phase of the filter, and the emission controller synchronizes the emission timing with the phase of the filter. 3. The radiation imaging apparatus of claim 1, wherein the filter includes multiple filter areas and the multiple filter areas are selectively inserted in sequence into the path. 4. The radiation imaging apparatus of claim 1, wherein the filter includes multiple filters each having at least one filter area, and the multiple filters are selectably used. 5. The radiation imaging apparatus of claim 4, wherein each of the multiple filters is individually rotatable, and arranged to be insertable into the path, and has a transmission area for retracting the at least one filter area of the filter from the path when the at least one filter area of another filter is inserted in the path. 6. A radiation imaging apparatus comprising:a radiation source for performing multiple radiation emissions to a subject;a filter having at least one filter area for changing energy distribution of radiation in at least one of the multiple radiation emissions, the filter periodically shifting between an inserted state in which the at least one filter area is inserted in a path of radiation and a retracted state in which the at least one filter area is retracted from the path;a cycle determining section for determining a shift cycle of the filter for shifting the at least one filter area between the inserted state and the retracted state based on subject information; anda driver for driving the filter to shift at the determined shift cycle, the driver driving the filter to start shifting before a start of the first radiation emission and to shift until an end of the last radiation emission after the first radiation emission, wherein the subject information is body thickness of the subject, and the cycle determining section determines the shift cycle in accordance with a maximum exposure time determined based on the body thickness. 7. An imaging control device for controlling a radiation source for performing multiple radiation emissions to a subject, and a filter having at least one filter area for changing energy distribution of radiation in at least one of the multiple radiation emissions, the filter periodically shifting between an inserted state in which the at least one filter area is inserted in a path of radiation and a retracted state in which the at least one filter area is retracted from the path, the imaging control device comprising:a cycle determining section for determining a shift cycle of the filter for shifting the at least one filter area between the inserted state and the retracted state before a start of the first radiation emission based on subject information;a controller for controlling a driver for driving the filter, the controller controlling the driver to shift the filter at the determined shift cycle by starting before a start of the first radiation emission and until an end of the last radiation emission after the first radiation emission; andan emission controller for outputting a signal for starting a radiation emission to the radiation source to control emission timing of radiation,wherein the subject information includes cardiac cycle information and respiration information of the subject, and the radiation imaging apparatus further includes a signal detector for detecting a cardiac signal indicating a state of the cardiac cycle information and a respiratory signal indicating a state of the respiration information, andwherein the cycle determining section determines the shift cycle so as to be in synchronization with the cardiac cycle information, and the emission controller synchronizes the emission timing with a phase of the respiration based on the respiratory signal. |
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summary | ||
description | The invention relates to a charged particle beam device for inspecting or structuring a specimen. In particular, the present invention relates to a focussing charged particle beam device comprising an aperture for shaping the aperture angle of the charged particle beam. Charged particle beam devices are becoming increasingly important for imaging and structuring micro- and nanometer sized structures and devices. While electron beams are preferred for imaging, ion beams are more suitable for machining a specimen, for example, by using the ion beam for etching, cutting or deposition. For inspecting or structuring a specimen efficiently with a high spatial resolution, it is important that the aperture of the charged particle beam device is well matched to the operational set up. For example, for obtaining minimum charged particle beam spot size, or a maximum beam current at a given beam spot size, the aperture has to be optimized with respect to aberrations of the lenses involved in the charged particle beam device, to diffraction which depends on the wavelength of the charged particles, to particle beam current which influences Coulomb interaction, and to system magnification. FIG. 1 illustrates schematically, as an example, a scanning charged particle beam device 1 having a charged particle beam source 5 that emits a charged particle beam 7, an extraction electrode 11 to accelerate the charged particles of the charged particle beam 7 to a desired beam energy, an aperture system 13 to define aperture angle α and beam current, and a focussing lens 9 to focus the charged particle beam onto a specimen 3. For completeness, FIG. 1 also depicts a scanning system 17 to scan the charged particle beam 7 across the surface of the specimen 3. The aperture system 13 of FIG. 1 depicts schematically three different circular apertures 13a, 13b and 13c which enable a person to operate the charged particle beam device at three different beam currents and aperture angles α. In the case of FIG. 1, the aperture angle α is defined by the maximum angle with respect to the optical axis 8 at which a ray of charged particles can pass through the opening 13a. Accordingly, the aperture angle α is defined by the diameter D of the opening 13a, and the distance L between the charged particle beam source 5 and the opening 13a The three different apertures 13a, 13b, 13c can be selected by using aperture drive 15 to linearly move one of the three apertures into the charged particle beam 7. The openings of the apertures 13a, 13b, 13c, of FIG. 1 are circular to provide that the respective aperture can be aligned to be fully rotational symmetric with respect to the optical axis 8. With full rotational symmetry, the aperture angle α is independent of the plane within which the aperture angle α is taken. Therefore, a fully rotationally symmetric aperture usually provides the highest focussing quality compared to systems with apertures of less rotational symmetry. However, the aperture system 13 of FIG. 1 with the three opening 13a, 13b, 13c, allows for only three different aperture angles α to optimize beam current and beam resolution. While it is true that aperture system 13 may be designed to have more than the three apertures of FIG. 1, the total number of apertures of a aperture system is always limited by tight space limitations and the constraint not to deter the electric field configuration within the beam column. Further, when shifting aperture system 13 to change from one aperture with a first diameter D to another aperture with a second diameter, beam operation is interrupted. Such interruptions make it difficult to adjust the aperture during operation. In addition, changing the aperture by shifting aperture system 13 requires each time an alignment procedure to align the new aperture to the optical beam axis. Such alignment procedure is generally time consuming. Further, permanent exposure of the aperture system 13 of FIG. 1 to a charged particle beam usually causes the aperture defining edges to change over time. For example, exposure to an electron beam generally leads to a contamination of the edges, while exposure to an ion beam generally leads to a removal of the aperture defining material. Both effects cause the aperture angle to drift over time which in turn causes beam spot size and beam current to vary uncontrollably. It is therefore a first aspect of the present invention to provide a charged particle beam device which does not show the above mentioned problems. It is yet a further aspect of the present invention to provide a charged particle beam device which provides more flexibility for adjusting the aperture size to optimize spatial resolution and beam current for any given application. It is yet a further aspect of the present invention to provide a charged particle beam device where focussing and beam current performance do not uncontrollably change due to deforming aperture shape or size induced by high beam exposure. Further advantages, features, aspects, and details of the invention are evident from the dependent claims, the description and the accompanying drawings. The claims are intended to be understood as a first non-limiting approach of defining the invention in general terms. The charged particle beam device according to claim 1 comprises a charged particle beam source to generate a charged particle beam, a focussing lens to focus the charged particle beam onto the specimen, and an aperture system which comprises a first member to block a first portion of the charged particle beam between the charged particle beam source and the focussing lens, a second member to block a second portion of the charged particle beam between the charged particle beam source and the focussing lens, first means for moving the first member to adjust the size of the blocked first portion of the charged particle beam, and second means for moving the second member independently of the first member. The present invention therefore is based on the idea to provide the aperture system with at least two independently movable members for defining an aperture angle for the charged particle beam device. This way, the aperture angle can be gradually increased by moving the at least two members in opposite directions away from the charged particle beam, and gradually decreased by moving the at least two members towards each other towards the charged particle beam. The fact that an aperture can be gradually increased or decreased enormously simplifies the search for an optimum aperture angle for a given application compared to the discrete selection of a limited number of apertures, as described in FIG. 1. Further, the adjustment of the aperture with the aperture system according to the invention can be carried out during beam operation, i.e. without having to interrupt the beam or even break the vacuum. This too greatly accelerates the options for operating and improving the performance of a charged particle beam device. With the capability of gradually adjusting the aperture angle and the alignment of the aperture with respect to the charged particle beam, position and size of an aperture can be optimized incrementally without having to interrupt probing or structuring of the specimen. This facilitates a fast and easy optimization of aperture size and position for any given application, e.g. for minimizing the size of the charged particle beam spot at a given beam current, or, vice versa, for maximizing the charged particle beam current at a given beam spot size. The invention therefore is also based on the idea to sacrifice the many advantages inherent in a fully circular aperture for an aperture system whose aperture is less circular but more flexible for defining size and alignment position for any given application. A further aspect of the present invention is that the movable members of the aperture system according to the invention can be moved with respect to the charged particle beam in a way that exposes different sections of the respective members to the beam without changing the shape of the aperture. This way, sections of the members that begin to deform or contaminate due to too much irradiation by the charged particle beam, can be exchanged during operation by new ones without that aperture changes. This way, the aperture system has a significant longer lifetime which reduces the down time of the charged particle beam device. A still further aspect of the present invention is that the means for moving a member can be used to align the aperture of the charged particle beam device to the optical axis, without affecting the aperture geometry. This alignment can be carried out during beam operation which makes it easy to determine a correct alignment. Such aperture system can spare the use of deflectors for alignment. The aperture system according to the invention includes at least two members to independently block a first and a second portion of a charged particle beam. With only two members, it is possible to provide a strip-like shaped cross section of the charged particle beam with a strip-width optimized to deliver a low spatial resolution in the direction of the strip-width and a high resolution in the direction perpendicular thereto. Having a high spatial resolution only within one direction may well be sufficient for applications where spatial resolution is required only within one dimension. In one preferred embodiment of the invention, the aperture system includes three members to independently block a first, second and third portion of a charged particle beam to define a triangular aperture that fully encircles the charged particle beam. This way, the aperture angles of the charged particle beam device can be fully controlled by the positioning of only three independently movable members. Preferably, the charged particle beam device also includes a magnetic or electric hexapole component to reduce the beam spot size by rounding the three corners of the triangular shaped beam spot obtained from passing the beam through the triangular aperture. In another preferred embodiment of the invention, the aperture system includes four members to independently block a first, second, third and fourth portion of a charged particle beam. Such aperture system is usually more complicated to build and operate than a system with only two members. However, with four members, it is possible to provide an adjustable rectangular aperture whose length and width can be optimized to deliver a high spatial resolution in two orthogonal dimensions. This makes it possible to deliver a small beam spot size for high spatial resolution in two dimensions. Preferably, the charged particle beam device with the aperture system of four members also includes a magnetic or electric octupole component. With the octupole component, it is possible to reduce the beam spot size by rounding the four corners of the rectangular shaped beam spot obtained from passing the beam through the aperture system. Since many charged particle beam devices are equipped with a magnetic or electric octupole component anyway, this method of focussing the charged particle beam is often easier to realize than installing and using eight movable members for defining the aperture. In another preferred embodiment of the invention, the charged particle beam device includes eight members to independently block a first, second, third, fourth, fifth, sixth, seventh and eighth portion of a charged particle beam. Such aperture system is usually more complicated to build and operate than a system with two or four members; however, with eight members, it is possible to provide an adjustable octagonal aperture whose eight sides can be optimized to deliver an even higher spatial resolution than a rectangular aperture can achieve. The term “charged particle beam device” in claim 1 refers to any device that uses a charged particle beam to probe or structure a specimen. Preferably, the charged particle beam device is a device for focussing the charged particle beam onto a specimen with a high spatial resolution. Preferably, the charged particle beam device includes a focussing lens for focussing an image of the charged particle beam source onto the specimen. Further, preferred, the charged particle beam device includes an aperture system having an aperture for defining an aperture angle at which the charged particle beam arrives at the focussing lens. Preferably, the aperture serves to limit the spherical and/or chromatic aberrations generated by the focussing lens. Further, depending on the application, a skilled person would know what other beam optical components to include to the charged particle beam device, like condensers, beam boosters, deflectors and the like. By imaging the charged particle beam source onto the specimen, the charged particle beam device can be used for applications that require the highest possible spatial resolution, e.g. from 1 micrometer down to 1 nanometer. For example, the charged particle beam device may be a scanning particle system for scanning a focussed charged particle beam across the specimen to inspect or structure the specimen. The charged particle beam devices may further be charged particle beam microscopes to probe a specimen, e.g. a scanning electron microscope (SEM), a transmission electron microscope (TEM), a scanning transmission microscope (STEM), or the like. Further, the charged particle beam device according to the invention may also be a device that uses the charged particle beam to structure a specimen. Non-limiting examples of charged particle beam devices that structure a specimen are, e.g. an electron beam pattern generators used to structure the surface of a specimen, like a lithographic mask, a focussing ion beam device (FIB) to slice or mill a specimen, and the like. Even though the term “charged particle beam” mainly refers to beams of electrons or beams of ions, the charged particle beam may also be of other charged elementary particles. The charged particle beam source according to the invention may be any source that is capable of emitting electrons, ions or other elementary particles into vacuum. Preferably, the charged particle beam source is one of the known electron beam sources used for electron microscopes, e.g. a thermionic tungsten hairpin gun, or one of the many types of field emission electron guns known in the art. If the charged particle beam device is an ion beam device, the charged particle beam source is preferably a Ga-ion beam source, or a gas plasma source. The term “focussing lens” according to the invention refers to any lens that is capable of providing a focussing electric or magnetic field for focussing the beam of charged particles like onto a specimen. The term “focussing lens” also includes lenses which combine electrical and magnetic fields for focussing the charged particle beam, see e.g. “High Precision electron optical system for absolute and CD-measurements on large specimens” by J. Frosien, S. Lanio, H. P. Feuerbaum, Nuclear Instruments and Methods in Physics Research A, 363 (1995) which herewith is included in the description. The term “aperture system” of the present invention refers to a system comprising at least a first member and a second member that are capable of blocking a respective first portion and/or second portion of the charged particle beam. Preferably, the members have a respective first and/or second edge that are capable of defining a respective first and/or second boundary of the aperture through which the charged particle beam can pass. Preferably, the term “aperture” refers to the area lateral to the charged particle beam direction left for the charged particle beam to pass by the members of the aperture system. If the members fully encircle the optical beam axis, the aperture may also be considered as the area lateral to the charged particle beam direction that is limited by the edges of the members surrounding the beam optical axis. Preferably, first and/or second members each comprise two facing surfaces which approach each other along at least one direction to form an edge. Preferably, first and/or second members have the shape of a blade of a knife. Preferably, first and/or second members are made of a conducting material like platinum or molybdenum in order to not charge up during blocking the first and/or second portion of the charged particle beam. Preferably, the first and/or second members are oriented in a way that the first and/or second edge face the charged particle beam like a knife blade intended to “cut through” the charged particle beam. Preferably, first and/or second edges are one-sided cut edges. In this case it is preferred that the cutting edge of the respective one-sided cut edge lies within the plane of the member surface facing the charged particle beam source. This way, the scattering of the charged particle beam at the edges of the members is minimized to provide sharp boundaries for the aperture. Further, preferably, first and/or second member is thick enough to absorb or back-scatter the incoming radiation of the charged particles. Preferably, first and/or second portions of the charged particle beam refer to the portions of the beam which is blocked by the respective member. The expression “blocked portion” refers to those particles of the charged particle beam that hit a member to become either absorbed or back-scattered by the member. The expression “transmitted portion”, in contrast, refers to those particles of the charged particle beam that are able to pass all members of the aperture system. Preferably, the edges of the respective first and second members are oriented with respect to each other as if they were to cut laterally through the charged particle beam from opposite sides. This way, the edges of the first and second members can limit the lateral extension of the charged particle beam to define an aperture angle for the focussing lens. Preferably, the edges of the respective first and second members are positioned to “cut through” the charged particle beam at the same location. Generally, it is preferred that the aperture system is designed to provide a high symmetry to make design, construction and operation of the charged particle beam device as easy as possible. It is therefore preferred that the shape of the first member is the same as the shape of the second member. Further, it is preferred that the first member can be moved into a position in which the first member and the second member are symmetrically aligned with respect to a rotation by 180 degrees around the optical axis of the charged particle beam device. It is further preferred that the first means for moving the first member and the second means for moving the second member are mechanically the same means. Preferably, the first edge and/or the second edge of the respective first and/or second member is shaped to provide a respective first and/or second boundary which extends essentially linearly. This way, by moving the respective member in a direction in which the respective edge extends, a new section of the edge can become exposed to the charged particle beam without that the shape of the effective aperture is changed. This way it is possible that sections of an edge that have been deformed due to interaction of the charged particle beam with the edge, can be replaced during operation by unused sections of the same edge without affecting the geometry of the aperture. Accordingly, time consuming replacements of outworn members, which requires breaking and renewing the vacuum of the charged particle beam device, can be saved. With members having a linearly extending edge, it is preferred that they are moved by means that are capable of moving the member independently in two different directions. Preferably, in this case, one direction would point towards the charged particle beam for defining the aperture angle of the charged particle beam and the other along the edge to “replace” the edge by replacing it with a new section of the edge, without changing the aperture geometry. In another preferred embodiment of the invention, first and/or second edges are shaped to provide inwardly angled respective first and/or second boundaries. This way, it is possible to provide with only two members a first and second boundary which fully encircles the charged particle beam. Preferably, the angle of the angled edges is essentially 90 degrees in order to provide a rectangular, possibly squared aperture. In still another preferred embodiment of the invention, the first and/or second edges are shaped to provide a rounded, preferably a circular first and/or second boundary. Preferably, first means and/or the second means for moving the respective first and/or second members are means that are capable of moving the respective member in steps having a step size smaller than 10 μm, preferably smaller than 1 μm. For some applications, it is preferred that the step size is even smaller than 0.1 μm. Preferably, first and/or second means each include a motor, preferably a step motor, or a piezo-drive in order to be able to move the respective member by such small step size. First and/or second means may also comprise encoding systems to improve precision and reproducibility for moving the respective members. In another preferred embodiment of the invention, the aperture system includes more than two members. This way, it is possible for the respective edges of the members to provide a boundaries that define an aperture that filly encircles the charged particle beam even if the edges of the more than two members all extend linearly. For example, with a first, a second and a third member having respective first, second and third linear edges, it is possible to form a triangular shaped aperture. Further, with a first, a second, a third and a fourth member having respective first, second, third and fourth linear edges, it is possible to form a rectangular, or even squared, aperture. Further, with a first, a second, a third, a fourth and a fifth member having respective first, second, third, fourth and fifth linear edges, it is possible to form a pentagonal shaped aperture, and so on. The more members the aperture system has, the higher the degree of symmetry with respect to a rotation around the charged particle beam axis (optical axis) can be achieved. It should be noted that a particularly preferred aperture system comprises eight members. With a first, a second, a third, a fourth, a fifth, a sixth, a seventh and an eighth member having respective first, second, third, fourth, fifth, sixth, seventh and eighth linear edges, it is possible to form an octagonal shaped aperture. Since the octagonal shape can be made to exhibit a higher degree of rotational symmetry with respect to the charged particle beam than a squared shaped aperture, an aperture system with eight members may significantly reduce the beam spot size, compared to the beam spot size shaped by, say, four members. Again, when having more than four members, it is generally preferred that the aperture system is capable of providing a high symmetry with respect to the charged particle beam in order to make design, construction and operation of the charged particle beam device as easy as possible. It is therefore preferred that the shape of all members involved in the aperture system are the same. Further, it is preferred that all members involved in the aperture system can be moved into positions that provide the highest possible rotational symmetry for the aperture. In the description of the preferred embodiments below, the reference numbers in the description refer to the enclosed figures FIG. 2, FIG. 3a-3d, 4a-4e, 5a-5b, and FIG. 6a-6b. However, the figures only represent particular, non-limiting embodiments of the invention which only have the purpose of being illustrative examples of the invention. The description below, even though it makes reference to the figures, is to be understood in a broad sense and includes any deviation from the described embodiments which is obvious to a person skilled in the art. FIG. 2 schematically illustrates relevant elements of a charged particle beam device 1 according to the invention, which may be a scanning electron microscope (SEM) of a type as described, for example, in the publication by J. Frosien et al. in “Nuclear Instruments and Methods in Physics Research”, A 363 (1995). In this case, the charged particle beam source 5 is a thermal field emission gun having a ZrO/W cathode in which the electron emission is initialized by an electric field between an emission tip 12 and extraction electrode 11. A further acceleration electrode 14 accelerates the emitted electrons to a desired intermediate beam energy of typically 5 to 20 keV. These energy values, however, may vary widely when other types of microscopes are used. Aperture system 13 between the charged particle beam source 5 and the focussing lens 9 is shown to block first and second portions 7a, 7b of the electron beam 7 thereby defining the shape of the aperture 6 and, accordingly, the aperture angle αx at which the electron beam 7 arrives at the focussing lens 9. The focussing lens 9 may be any of the known electrostatic lenses, magnetic lenses or, as in FIG. 2, combined electrostatic magnetic lenses which are described in more detail in the mentioned publication by J. Frosien. The combined electrostatic magnetic lens of FIG. 2 includes a magnetic lens 19 where the yoke is also used as an electrode to provide an electric field. For more completeness, FIG. 2 also depicts a scanning system 17 to scan the charged particle beam 7 across the surface of the specimen 3. FIG. 2 schematically illustrates the relevant elements of the aperture system 13. The aperture system 13 includes two members, i.e. first member 20 having a first edge 22 and second member 30 having a second edge 32, to define the size of the aperture 6 and the aperture angle αx within the plane defined by the optical axis 8 and the X-direction. The two edges 22, 32 correspond to the sharp side of the respective member 20, 30 which in general has a knife-like shape. This is to minimize scattering of electrons of the electron beam 7 grazing along any of the edges 22, 32. Preferably, the two members are positioned to “cut” into the charged particle beam 7 at the same height (i.e. Z-direction) to provide that the edges 22, 32 block the respective first and second portions 7a, 7b of the electron beam 7 at the same location. First and second member 20, 30 are each mounted to respective first and second means 24, 34 for moving the members, which in turn are fastened to respective stages 26, 36 that are part of the charged particle beam device structure. With first and second means 24, 34, the respective members 20, 30 can be moved with respect to the first and second stages 26, 36. For moving, first and second means 24, 34 each comprise two piezo-drives 24a, 24b, 34a, 34b. Like for a X-Y cross table, piezo-drives 24a, 34aserve to move the respective members 20, 30 in X-direction for moving the respective edges 22, 32 in or out of the electron beam 7, while piezo-drives 24b, 34b serve to move the respective members 20, 30 in a Y-direction to move the respective edges 22, 32 in a direction along the direction of the respective edge (see also FIG. 3a). This way, piezo-drives 24a, 34a can be used to adjust the aperture angle αx to a desired angle, while piezo-drives 24b, 34b can independently be used to “clean” the edges 22, 32 by moving one section of a given edge 22, 32 out of the electron beam 7 out and another section in. Note that the “cleaning” -procedure can be carried out without changing the aperture. The piezo-drives 24a, 34a, 24b, 34bin FIG. 2 are capable of moving the members with step sizes smaller than 10 μm. Generally, the use of piezodrives for X-Y cross tables is well known in the art and does not require further explanation. It should be noted that the present invention is independent of the means by which the members are moved. Therefore, besides using piezo-drives as means for moving a member, the members can be moved, e.g., by stepping motors, by thermal expanding materials or memory metals that can move a member, or by any other means that are capable of moving a member within the micrometer scale. Also, the means for moving a member may include an encoder to improve precision and reproducibility for the positioning of the respective members. FIG. 3a illustrates the aperture system 13 of FIG. 2 as seen in the direction of the electron beam 7, whose cross section at the aperture system 13 is indicated by the dotted circle. Typically, the diameters of electron beams in an SEM of the type shown in FIG. 2 at the position of the aperture are in the range of 1 μm to 500 μm and preferably in the range between 5 and 200 μm. First and second members 20, 30 of FIG. 3a are equally shaped and have each linearly extending edges 22, 32 that run in parallel thereby defining a slit 16 of width D. At the same time, the two edges 22, 32 define a first boundary 28 and a second boundary 38 that define the shape of aperture 6. Aperture 6 of FIG. 3a is characterized by the width D of the slit 16 which in turn defines an aperture angle αx in X-direction. The aperture angle αy in Y-direction, in contrast, is only limited by the diameter of the electron beam 7. The sizes of the members 20, 30 are limited by the available space within the vacuum chamber of the respective charged particle beam device. On the other hand, the size of the members must be large enough to be able to block the relevant portions 7a, 7b of the electron beam 7. Further, it is advantageous to have the linearly extending first and second edges 22, 32 at least a few times longer than the cross section of the electron beam 7 in order to be able to “clean” the edges by replacing a worn out section of the edge by a new section that has not been exposed to the electron beam 7. Therefore, for a electron beam device as shown in FIG. 2, the sizes of the members 20, 30 in X- and Y-directions are typically in the range of a few millimeters. FIG. 3a further illustrates schematically, as dotted lines, first and second means 24, 34 sandwiched between the respective first and second stages 26, 36 and the respective first and second members 20, 30. As already explained in the description of FIG. 2, first and second means 24, 34 each comprise piezo-drives 24a, 34a for driving the respective members 20, 30 in X-direction, and piezo-drives 24b, 34b for driving the respective members 20, 30 in Y-direction, as indicated by the four double arrows in the FIG. 3a. FIG. 3b illustrates a second aperture system 13 according to the invention that, for example in the SEM of FIG. 2, can serve as a replacement for the aperture system 13 of FIG. 3a. The aperture system of FIG. 3b is the same as the one of FIG. 3a with the differences that the first and second edges 22, 32 are inwardly angled in order to provide for an aperture 6 that fully encircles the transmitted portion of the electron beam 7. The respective angles of the angled edges 22, 32 in FIG. 3b are both essentially 90 degrees in order to have a rectangular shaped or, preferably, a square shaped aperture 6. With the rectangular edged members 20, 30, it is possible to fully control the aperture angles of the charged particle beam device in any direction within the X-Y-plane with only two members 20, 30. For example, by moving first and second members 20, 30 apart from each other along the X-direction, the aperture 6 can be gradually increased while maintaining the proportions of the two sides of the aperture 6. However, with the design of FIG. 3b, it is not possible to “clean” any of the edges 22, 32 by simply moving the members 20, 30 in a direction, as it was possible with the linearly edged aperture system 13 of FIG. 3a. The disadvantage of not being able to “clean” the edges 22, 32 of the members 20, 30 of FIG. 3b within the vacuum can be remedied by the “saw tooth” design of the two members 20, 30 shown in FIG. 3c. The two members 20, 30 of FIG. 3c exhibit each multiple inwardly angled edges 22, 32 which together resemble a line of teeth of a saw. The two members 20 30 with the saw tooth sides facing each other can be positioned with respect to each other to provide multiple apertures 6, 6a, 6b. For example, the “active” aperture defining the aperture angle of the charged particle beam 7 in FIG. 3c is aperture 6. However, once the first and second edges 22, 32 at the regions of the respective boundaries 28, 38 are worn out due to irradiation by the charged particle beam 7, the active aperture 6 can be replaced by one of the other apertures 6a or 6b by moving the first and second members 20, 30 in into the same Y-direction. The moving of the two members, again, can be carried out by means of the first and second means 24, 34 for moving the respective first and second members 20, 30, e.g. a piezo-drive, without having to break the vacuum. FIG. 3d illustrates a fourth aperture system 13 according to the invention that, for example in the SEM of FIG. 2, can serve as a replacement for one of the aperture systems 13 of FIGS. 3a-3cb. Aperture system 13 of FIG. 3d comprises four members, i.e. first member 20, second member 30, third member 40 and fourth member 50, in order to provide four boundaries 28, 38, 48 and 58 that can be positioned to fully encircle the electron beam 7. In addition, maximum rotational symmetry of the transmitted portion 6 with respect to the electron beam 7 is achieved by having the first and second edges 22, 32 aligned in parallel, the third and fourth edges 42, 52 aligned in parallel and by having the first or second edge 22, 32 orthogonally aligned with respect to the third and fourth edges 42, 52 (see also FIG. 4b). This way, with the help of means 24, 34, 44, 54 for moving first, second, third or fourth members 20, 30, 40, 50, the members can be moved to positions that provide a square shaped aperture 6 aligned to the electron beam. Again, for the sake of simplicity in manufacturing and operation of a charged particle beam device 1 according to the invention, it is preferred for the aperture system 13 of FIG. 3d that the first member 20, first means 24 for moving the member and first stage 26 are essentially the same units as the corresponding second, third and fourth member 30, 40, 50, the corresponding second, third and fourth means 34, 44, 54 for moving the respective member, and the corresponding second, third and fourth stage 36,46, 56. Further, for symmetry reasons, it is preferred that the aperture system 13 of FIG. 3e can be considered as being assembled by two aperture systems of the type shown in FIG. 3a, with are positioned on top of each other and rotated by 90 degrees with respect to each other within the X-Y-plane. In this case, the first aperture system with first and second members 20, 30 blocks first and second portions 7a, 7b of the electron beam 7, while the second aperture system with third and fourth members 40, 50 blocks third and fourth portions 7d, 7e. In order to illustrate the advantages of the aperture systems according to the invention, FIGS. 4a-4e use the aperture system 13 as described in FIG. 3d to demonstrate several operational modes. FIG. 4a deals with the situation where the first, second, third and fourth edges 22, 32, 42, 52 are not centered with respect to the optical axis 8 of the charged particle beam device. In this case, aperture 6 is not aligned with respect to the optical axis 8. This often implies that the charged particle beam 7 becomes deformed due to aberrations when passing through the focussing lens 9. Such deformation in turn usually reduces the spatial resolution and beam current for probing or structuring a specimen. However, due to the ability of the aperture system 13 to move the members 20, 30, 40 and 50 gradually and independently of each other, the alignment of the first, second, third and fourth edges 22, 32, 42, 52 with respect to the optical axis 8 can be recovered, e.g. by moving second member 30 to the left within the X-direction, and by moving the fourth member 50 upwards within the Y-direction. After such operations, the transmitted portion 6 is aligned with respect to the optical axis, as shown in FIG. 4b. FIG. 4c illustrates, with four arrows, movements of the first, second, third and fourth members 20, 30 40 50 in order to increase the aperture 6 in X- and in Y-directions without affecting its alignment to the optical axis 8. Starting out from a situation as shown in FIG. 4b, first member 20 is moved along the X-direction to the left by a distance d, second member 30 is moved along the X-direction to the right by the same distance d, third member 40 is moved along the Y-direction upwards by the same distance d, and fourth member 50 is moved along the Y-direction downwards by the same distance d. FIG. 4d illustrates, by means of arrows, movements of the first, second, third and fourth members 20, 30, 40, 50 in order to “clean” aperture 6. “Cleaning” is carried out by replacing deformed or contaminated sections of respective first, second, third and fourth edges 22, 32, 42, 52 by sections of the same edge which have not been exposed to the charged particle beam 7 before. Deformation or contamination of a section of an edge is usually caused by the intense exposure to the charged particle beam 7 during the blocking of portions 7a, 7b, 7d, 7e of the charged particle beam to define the shape of the aperture 6. The replacement of deformed or contaminated sections of an edge by sections which have not been exposed to the charged particle beam in the way as described in FIG. 4d makes it possible to maintain a precise definition of the aperture 6 and the corresponding aperture angles over long periods of time. In the design of FIG. 4d, the replacement of the deformed or contaminated sections is carried out by moving the respective members 20, 30, 40, 50 in a direction parallel to the respective edge 22, 32, 42, 52. For example, in FIG. 4d, the first member 20 is moved downwards along the Y-direction for at least the distance corresponding to the diameter of the electron beam 7, the second member 30 is moved upwards along the Y-direction for at least the distance corresponding to the diameter of the electron beam 7, and the third and fourth members 40, 50 are moved to the left along the X-direction for at least the distance corresponding to the diameter of the electron beam 7. This way, the deformed or contaminated sections of all for members 20, 30, 40, 50 have been replaced by new edge sections without having changed the position or size of the aperture 6. Since the replacements of deformed or contaminated sections of edges by new edges can be carried without braking the vacuum of the charged particle beam device, or even during beam operation, the working efficiency of charged particle beam devices can be significantly improved. FIG. 4e illustrates the situation where the first and second members 20, 30 have been moved towards each other along the X-direction in order to provide a small aperture angle ax within the X-direction, and where the third and fourth members 40, 50 have been moved in opposite directions along the Y-direction in order to make the aperture angle αy within the Y-direction large. This way, the electron beam 7 receives a rectangular shaped cross section after it has passed through the aperture 6 defined by the first, second, third and fourth edges 22, 32, 42, 52. Such shape can be useful for beam operations where a high spatial resolution is required only in one direction, like for ion beam (FIB) devices used to mill slices out of a specimen. For such operation, it is even conceivable to place the four stages 26, 36, 46, 56 onto a rotational table in order to freely rotate the aperture for efficiently milling the specimen along any desired direction. FIGS. 5a and 5b illustrate a fourth aperture system 13 according to the invention using eight members, i.e. the first member 20, second member 30, third member 40, fourth member 50, fifth member 60, sixth member 70, seventh member 80 and eighth member 90, to define an octagonal shaped aperture 6. With eight members 20, 30, 40, 50, 60, 70, 80, 90, it is possible to “cut away” the four corners of a rectangular shaped aperture 6 that an aperture system with only four members would produce. With the corners “cut off”, the rotational symmetry of the aperture 6 further improves to minimize the beam spot size. The aperture system 13 of FIG. 5a essentially consists of two aperture systems of the type shown in FIG. 3d which are position on top of each other and rotated by 45 degrees with respect to each other within the X-Y-plane. In this case, the first aperture system comprises the first, second, third and fourth members 20, 30, 40, 50, while the second system comprises the fifth, sixth, seventh and eighth members 60, 70, 80, 90. Each of the first, second, third, fourth, fifth, sixth, seventh and eighth members 20, 30, 40, 50, 60, 70, 80, 90 has respective linear first, second, third, fourth, fifth, sixth, seventh and eighth edges 22, 32, 42, 52, 62, 72, 82, 92 of the type as described for FIG. 3d. Further, first, second, third, fourth, fifth, sixth, seventh and eighth members 20, 30, 40, 50, 60, 70, 80, 90 are each mounted to respective independent means for moving the respective members (not shown in FIG. 5a) with respect to the first, second, third, fourth, fifth, sixth, seventh and eighth stages 26, 36, 46, 56, 66, 76, 86, 96. FIG. 5b represents a blow-up view of the aperture region of the aperture system of FIG. 5a where the charged particle beam (not shown in FIG. 5a, b) is to arrive at the aperture system. FIG. The shape of the aperture 6 is defined by the first, second, third, fourth, fifth, sixth, seventh and eighth boundaries 28, 38, 48, 58, 68, 78, 88, 98 defined by the first, second, third, fourth, fifth, sixth, seventh and eighth edges 22, 32, 42, 52, 62, 72, 82, 92 which also define the portions of the charged particle beam 7 that are blocked by the first, second, third, fourth, fifth, sixth, seventh and eighth members 20, 30, 40, 50, 60, 70, 80, 90. FIGS. 6a illustrates a second charged particle beam device 1 according to the invention which equals the SEM of FIG. 2 in many ways. However, different from the SEM of FIG. 2, the charged particle beam device 1 of FIG. 6a has an aperture system 13 with four members 20, 30, 40, 50 like the one described in FIG. 3d. With the aperture system 13 of FIG. 3d, the charged particle beam device 1 is capable of producing an electron beam with a square-shaped cross section that results in a square shaped beam spot 120. In addition to the aperture system 13 of FIG. 3d, the charged particle beam device 1 of FIG. 6a also includes a magnetic octupole 101 positioned into the charged particle beam 7. As it turns out, the magnetic octupole field of the magnetic octupole 13 can be used to deflect the electrons of the charged particle beam 7 in a way that leads to a rounding of the four corners of the square shaped beam spot 120, which in turn reduces the size of the beam spot. The magnetic octupole 101, in principle, may be positioned anywhere within the charged particle beam 7 to provide for a reduction of the beam spot size. However, it is generally preferred that the magnetic octupole is positioned in regions of the charged particle beam 7 where the beam has a large cross section. Therefore, it is preferred that the magnetic octupole 101 is positioned in the region between the aperture system 13 and specimen 3 or, even better, nearby or within the focussing lens 9. FIGS. 6b and 6c illustrate the effect of using a magnetic octupole for rounding the corners of a squared shaped beam spot 120. In FIG. 6b, the magnetic octupole field is switched off. Accordingly, the beam spot 120 of the electron beam 7 after it has passed through the square-shaped aperture 13 and focussing lens 9, is square-shaped. The maximum extension of the square-shaped beam spot 120, therefore, is given by the diagonal of the square. FIG. 6c depicts the beam spot 120 of the electron beam 7 under the same conditions as in FIG. 6b, but with a magnetic octupole field switched on. The magnetic octupole field has the effect that charged particles in the corners of the charged particle beam 7 become deflected towards the optical axis with the effect that the corners of the beam spot 120 disappear. In this case, the maximum extension of the rounded beam spot 120 is essentially given by the length of a side of the original square, which is smaller than the diagonal of the square. As a consequence, the charged particle beam 7 corrected by the magnetic octupole 101 in FIG. 6a can generate a smaller beam spot with a higher current density. The method of reducing the spot size of a square-shaped charged particle beam 7 by a magnetic or electric octupole has the advantage over the aperture system 13 with octagonal aperture (see FIG. 5b) in that many charged particle beam devices are equipped with a magnetic or electric octupole anyway in order to correct beam astigmatism. Therefore, instead of installing another four movably connected members of the type as shown in FIG. 5b into a charged particle beam device, it is simpler to use the octupole field of a stigmator for correcting the square-shaped beam spot 120 to a rounded beam spot. FIGS. 7a-c demonstrate a focussing concept similar to the one of FIG. 6a-c, with the difference that the aperture 6 of the charged particle beam device is formed of three edges 22, 32, 42, instead of four, and that instead of a magnetic octupole 101, a magnetic or electric hexapole is used. FIG. 7a shows the aperture system 13 with the members 20, 30, 40 moved to form a triangular shaped aperture 6 and, in addition, aligned to the optical axis 8 of the charged particle beam 7. Preferably, first, second and/or third edges 22, 32, 33 of the first, second and third members 20, 30, 40 are essentially linear. This way, it is possible to move the members 20, 30, 40 in the direction of the respective edges 22, 32, 42 without changing the aperture shape. This allows for a “cleaning” of the aperture boundaries 28, 38, 48 without changing the shape of the aperture 6. Further, it is preferred that first, second and third members 20, 30, 40 are oriented in a way that the respective edges 22, 32, 42 form an aperture 6 of essentially equilateral triangled shape. This helps to maximize the rotational symmetry with respect to the beam optical axis. In the example of FIG. 7a, first member 20, in the projection along the beam axis, is located between second and third members 30, 40 with the second member closest to the charged particle beam source. Further, each member 20, 30, 40 is independently movably connected to the respective stages 26, 36, 46 in two orthogonal directions in order to adjust the size of the aperture 6, align the aperture 6 to the charged particle beam 7 and to “clean” the edges. With a triangular shaped aperture 6, the beam spot 120 of the charged particle beam is shaped like a triangle, as shown in FIG. 7b. The size of the beam spot can 120 further be reduced if a magnetic and/or electric hexapole field generated by a magnetic and/or electric hexapole is applied to the charged particle beam 7. As it turns out, the magnetic and/or electric hexapole field is capable of deflecting the electrons of the charged particle beam 7 in a way that leads to a rounding of the three corners of the triangular shaped beam spot 120, which in turn reduces the size of the beam spot. The magnetic and/or electric hexapole, like the magnetic octupole 101 of FIG. 6a, may be positioned anywhere within the charged particle beam 7 to provide for a reduction of the beam spot size. However, it is generally preferred that the hexapole is positioned in regions of the charged particle beam 7 where the beam has a large cross section. Therefore, it is preferred that the magnetic and/or electric hexapole is positioned in the region between the aperture system 13 and specimen 3 or, even better, nearby or within the focussing lens 9. FIGS. 7b and 7c show the beam spots 120 with the aperture system 13 of FIG. 7a with a magnetic or electric hexapole component switched off (FIG. 7b) and switched on (FIG. 7c). As expected, the beam spot 120 with the hexapole component switched off has the shape of aperture 6 of FIG. 7a, i.e. it is triangular. With the hexapole component switched on, however, the three corners of the beam spot 120 of FIG. 7b become rounded which in turn reduces the size of the beam spot 120. This further improves the spatial resolution of the charged particle beam device for scanning or structuring a specimen. |
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044477309 | abstract | A transportation container for radioactive materials, especially for irradiated fuel elements, has a cylindrical body of a material to shield gammaradiation. The exterior of the body is provided with spaced cooling ribs. In order to avoid the continuation of the crack into the container body occasioned by damage to the cooling ribs, as by a fall, the body is provided with spaced flanges, of lesser height than the ribs, extending transversely of and to and between the ribs so that any cracking or breaking off of a cooling rib will not occur at its base. |
abstract | An electro-technical device includes a first housing portion electrically isolated from a second housing portion with a point source being disposed within the first housing portion. A movable conductor is connected to the first portion and is responsive to an electric field generated by the point source to cause the movable conductor to contact the second housing portion to complete a circuit and send out a control signal. |
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047599120 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like are words of convenience and are not to be construed as limiting terms. In General Referring now to the drawings, and particularly to FIGS. 1 to 5, there is shown a nuclear fuel assembly, generally designated 10, for a BWR to which the improved features of the present invention are advantageously applied. The fuel assembly 10 includes an elongated outer tubular flow channel 12 that extends along substantially the entire length of the fuel assembly 10 and interconnects an upper support fixture or top nozzle 14 with a lower base or bottom nozzle 16. The bottom nozzle 16 which serves as an inlet for coolant flow into the outer channel 12 of the fuel assembly 10 includes a plurality of legs 18 for guiding the bottom nozzle 16 and the fuel assembly 10 into a reactor core support plate (not shown) or into fuel storage racks, for example in a spent fuel pool. The outer flow channel 12 generally of rectangular cross-section is made up of four interconnected vertical walls 20 each being displaced about ninety degrees one from the next. Formed in a spaced apart relationship in, and extending in a vertical row at a central location along, the inner surface of each wall 20 of the outer flow channel 12, is a plurality of structural ribs 22. The outer flow channel 12, and thus the ribs 22 formed therein, are preferably formed of a metal material, such as an alloy of zirconium, commonly referred to as Zircaloy. Above the upper ends of the structural ribs 22, a plurality of upwardly-extending attachment studs 24 fixed on the walls 20 of the outer flow channel 12 are used to interconnect the top nozzle 14 to the channel 12. For improving neutron moderation and economy, a hollow water cross, generally designated 26, extends axially through the outer channel 12 so as to provide an open inner channel 28 for subcooled moderator flow through the fuel assembly 10 and to divide the fuel assembly into four, separate, elongated compartments 30. The water cross 26 has a plurality of four radial panels 32 composed by a plurality of four, elongated, generally L-shaped, metal angles or sheet members 34 that extend generally along the entire length of the channel 12 and are interconnected and spaced apart by a series of elements in the form of dimples 36 formed in the sheet members 34 of each panel 32 and extending therebetween. The dimpels 36 are provided in opposiong pairs that contact each other along the lengths of the sheet members 34 to maintain the facing portions of the members in a proper spaced-apart relationship. The pairs of contacting dimples 36 are connected together such as by welding to ensure that the spacing between the sheet members 34 forming the panels 32 of the central water cross 26 is accurately maintained. The hollow water cross 26 is mounted to the angularly-displaced walls 20 of the outer channel 12. Preferably, the outer, elongated longitudinal edges 38 of the panels 32 of the water cross 26 are connected such as by welding to the structural ribs 22 along the lengths thereof in order to securely retain the water cross 26 in its desired central position within the fuel assembly 10. Further, the inner ends of the panels together with the outer ends thereof define the inner central cruciform channel 28 which extends the axial length of the hollow water cross 26. Also, the water cross 26 has a lower flow inlet end 39 and an opposite upper flow outlet end 40 which each communicate with the inner channel 28 for providing subcoolant flow therethrough. Disposed within the channel 12 is a bundle of fuel rods 42 which, in the illustrated embodiment, number sixty-four and form an 8.times.8 array. The fuel rod bundle is, in turn, separated into four mini-bundles thereof by the water cross 26. The fuel rods 42 of each mini-bundle, such being sixteen in number in a 4.times.4 array, extend in laterally spaced apart relationship between an upper tie plate 44 and a lower tie plate 46 and connected together with the tie plates comprise a separate fuel rod subassembly 48 within each of the compartments 30 of the channel 12. A plurality of grids or spacers 50 axially spaced along the fuel rods 42 of each fuel rod subassembly 48 maintain the fuel rods in their laterally spaced relationships. Coolant flow paths and cross-flow communication are provided between the fuel rod subassemblies 48 in the respective separate compartments 30 of the fuel assembly 10 by a plurality of openings 52 formed between each of the structural ribs 22 along the lengths thereof. Coolant flow through the openings 52 serves to equalize the hydraulic pressure between the four separate compartments 30, thereby minimizing the possibility of thermal hydrodynamic instability between the separate fuel rod subassemblies 48. The above-described basic components of the BWR fuel assembly 10 are known in the prior art, except for modification of the Fuel rods 42 to comprise part of the hybrid fuel design of the present invention, as described hereinafter. The BWR fuel assembly 10, disclosed in greater detail in the patent to Barry et al cited above, has been discussed in sufficient detail herein to enable one skilled in the art to understand the hybrid fuel design of the present invention presented hereinafter. For a more detailed description of the construction of the BWR fuel assembly, attention is directed to the above-mentioned Barry et al patent. Hybrid Fuel Design for Avoiding PCI Constraints and Failures The present invention provides improved features in the form of a hybrid fuel design which minimizes PCI constraints and eliminates PCI failures. Referring to FIGS. 4-7, the hybrid fuel design comprises a cladded rod-type nuclear fuel in the form of the fuel rods 42 subdivided into the mini-bundles thereof being located within the compartments 30 and spaced transversely from both the walls 20 of the outer channel 12 and the radial panels 32 of the water cross 26, and a cladded plate-type nuclear fuel in the form of elongated fuel plates 54 respectively attached on the exterior of the radial panels 32 of the water cross 26. Power generation in the fuel assembly 10 is now distributed between the fuel rods 42 and fuel plates 54 such that there is reduced power generation per fuel rod/plate which minimizes PCI failures and avoids any nuclear/hydraulic stability problems. For example, power generation per fuel rod has been reduced approximately twenty percent from what it was heretofore, based on the twenty-five percent increase in heated perimeter. More particularly, each fuel plate 54 attached on the exterior of a water cross radial panel 32 is, preferably, generally coextensive in length with the fuel rods 42. Also, preferably the fuel plate 54 is generally coextensive in width with the water cross panel 32 but shorter in length than the panel. Additionally, each fuel plate 54 includes an arcuate-shaped (right-angled) inner sheet 56 of nuclear fuel disposed adjacent the exterior of each of a pair of the water cross panels 32 and an arcuate-shaped (right-angled) outer sheet 58 of cladding disposed adjacent the exterior of the inner sheet 56. The outer sheet at its periphery 60 has an inwardly-turned rim 62 by which it is attached to the pair of water cross panels 32 so as to sealably encase or enclose the inner sheet 56 of nuclear fuel from contact with the flowing coolant/moderator liquid. The nuclear fuel of the inner sheet 56 can be the same composition as the nuclear fuel in pellet form in the fuel rods 42, for example, uranium oxide. The cladding of the outer sheet 58 can be the same composition as the cladding of the fuel rods 42, such as zirconium oxide, which also can be the composition of the water cross 26. It is thought that the invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
046630856 | claims | 1. An apparatus for the decontamination of a metallic object contaminated by radiation, comprising: an electrolytic cell containing an electrolytic solution formed of an aqueous cerium nitrate solution containing trivalent cerium ions and tetravelent cerium ions, an anode and a cathode immersed in said electrolytic solution in said electrolytic cell and connected to a direct current power source, a feed pipe connected to a lateral side of said electrolytic cell and including a liquid pump, a decontamination cell connected to said feed pipe through said liquid pump, said decontamination cell containing said radiation-contaminated metallic object, a drain line connected from a lower section of said decomtamination cell to said electrolytic cell, said drain line including a filter and a circulation pump, a waste gas pipe connected to upper sections of said electrolytic cell and said decontamination cell to remove gases from said cells, a condenser connected to said waste gas pipe to liquefy said gases removed from said cells, a return pipe connected from said condenser to said electrolytic cell to return said liquefied gases to said electrolytic cells, a gas outlet pipe for leading out waste gas emanating from said condenser, and a demister and a waste gas blower connected to said gas outlet and adapted to recover water-nitric acid vapor and mist from said waste gas. an electrolytic cell containing an electrolytic solution formed of an aqueous cerium nitrate solution formed of trivalent cerium ions and tetravalent cerium ions, an anode and a cathode immersed in said electrolytic solution and connected to a direct current power source, a drain tank connected to said electrolytic cell through an overflow pipe, a first pipe connected through a valve to said drain tank, a decontamination cell holding said radiation-contaminated metallic object, a feed pipe connected to said decontamination cell through a liquid pipe and connected to said first pipe through a first pump, a rinse water tank connected through a rinse water return pipe to a lower section of said decontamination cell, a second pump connected through a pump to a lower section of said rinse water tank and through a third pipe to said drain tank, said liquid pipe being connected through a feed pipe for electrolysis to said pump and to said electrolytic cell, waste gas pipes connected one each to upper sections of said electrolytic cell, said decontamination cell, and said drain tank, a waste gas duct for collectively leading out waste gas from said waste gas pipes, a nitric acid vapor recovery device connected to said waste gas duct, a receptacle connected to said nitric acid vapor recovery device, a mist recovery device connected through a waste gas pipe to the empty space in an upper section of said receptacle, a waste gas blower connected through a waste gas pipe to said mist recovery device, and a recovery return pipe connected from the lower section of said receptacle to said drain tank. an electrolytic solution cell holding an electrolytic solution formed of an aqueous cerium nitrite solution containing trivalent cerium ions and tetravalent cerium ions, said electrolytic solution cell including first and second cell sections; an anode and a cathode immersed in said electrolytic solution and connected to a direct current power source, said anode and cathode being contained in said first electrolytic cell section; a contaminated metallic object holder for holding a radiation-contaminated metallic object immersed in said electrolytic solution, said object holder being contained in said second electrolytic cell section; a circulation pipe connected between said first cell section and said second cell section of said electrolytic solution cell for allowing said electrolytic solution in circulate; and a filter and a circulation pump coupled to said circulation pipe to cause said electrolytic solution to circulate in said pipe. 2. An apparatus according to claim 1, wherein said electrolytic cell is provided with a heater for heating said electrolytic solution. 3. An apparatus according to claim 1, wherein a branched pipe is disposed on the downstream side of said feed pipe, wherein a leading end of said branched pipe and a leading end of said feed pipe are connected to annular pipes disposed inside said decontamination cell. 4. An apparatus according to claim 3, wherein said annular pipes are disposed stepwise within said decontamination cell and slanted nozzles are connected one each to said annular pipes. 5. An apparatus according to claim 1, wherein a performated plate containing a through hole at the center thereof is disposed, through a support valve, in the lower section of said decontamination cell. 6. An apparatus for the decontamination of a metallic object contaminated by radiation, comprising: 7. An apparatus according to claim 6, wherein a filter is disposed between said feed pipe for electrolysis and said liquid pipe and said filter is enclosed with a radiation shielding member. 8. An apparatus according to claim 6, wherein a filter is disposed between said feed pipe for decontamination and said liquid pipe and said filter is enclosed with a radiation shielding member. 9. An apparatus according to claim 6, wherein said electrolytic cell and said decontamination cell are each provided with a heater for heating said electrolytic solution. 10. An apparatus according to claim 6, wherein a feedback pipe for spraying is connected through a pipe to said mist recovery device and a spray nozzle is connected to said feedback pipe so as to effect spraying of a second filter within said recovery device. 11. An apparatus according to claim 7 or claim 9, wherein pressure gauges are incorporated one each in pipes on the upstream side and downstream side of said filter. 12. An apparatus for the decontamination of a metallic object contaminated by radiation, comprising: 13. An apparatus according to claim 12, wherein said first cell section and said second cell section each have a sealed empty space in the upper portion thereof connected to a condenser, a demister and a waste gas blower through a waste gas pipe means. |
059603682 | abstract | The present invention is directed to a process for reducing the volume of low level radioactive and mixed waste to enable the waste to be more economically stored in a suitable repository, and for placing the waste into a form suitable for permanent disposal. The invention involves a process for preparing radioactive, hazardous, or mixed waste for storage by contacting the waste starting material containing at least one organic carbon-containing compound and at least one radioactive or hazardous waste component with nitric acid and phosphoric acid simultaneously at a contacting temperature in the range of about 140.degree. C. to about 210 .degree. C. for a period of time sufficient to oxidize at least a portion of the organic carbon-containing compound to gaseous products, thereby producing a residual concentrated waste product containing substantially all of said radioactive or inorganic hazardous waste component; and immobilizing the residual concentrated waste product in a solid phosphate-based ceramic or glass form. |
052710530 | abstract | A holddown leaf spring for a nuclear fuel assembly (10) with an upper end fitting (12) having spring retaining slots (30). Two spring stacks (20, 24 and 22, 26), respectively, are made up of unitary elongated metal bar springs having two substantially tapered width leg portions (20a, 20b, etc.) joined by an arcuate transition portion (28) therebetween. The wider opposite end portions (20c, 20d, etc.) are received in slots (30) of end fitting (12) and retained by welded pins (or capped screws) (not shown). The tapered width distributes localized stresses and may be horizontal or vertical. Less chance of debris exists due to retained spring ends. Flow at two opposite corners is opened up by the taper and by transition portion (28) being spaced from the nozzle openings (32 and 34). The assembly requires fewer parts and less machining. |
054770530 | claims | 1. A radiographic intensifying screen which comprises a support, a fluorescent layer formed on the support, a water repellent layer provided on the fluorescent layer and a protective layer formed by coating a solution containing a protective layer-forming resin on the water repellent layer. 2. The radiographic intensifying screen according to claim 1, wherein the fluorescent layer contains a binder in an amount of 1 to 10 parts by weight per 100 parts by weight of a phosphor. 3. The radiographic intensifying screen according to claim 1, wherein the water repellent is an organic silicon compound. 4. The radiographic intensifying screen according to claim 3, wherein the organic silicon compound is an alkylalkoxysilane. 5. A radiographic intensifying screen which comprises a support, a fluorescent layer formed on the support, and a protective layer formed by coating a solution containing a protective layer-forming resin on the fluorescent layer, wherein the fluorescent layer contains a water repellent, and wherein the water repellent is distributed so that the proportion of the water repellent in the fluorescent layer near the interface to the protective layer is higher than the proportion of the water repellent in the fluorescent layer near the interface to the support. 6. The radiographic intensifying screen according to claim 5, wherein the fluorescent layer contains a binder in an amount of 1 to 10 parts by weight per 100 parts by weight of a phosphor. 7. The radiographic intensifying screen according to claim 5, wherein the water repellent is an organic silicon compound. 8. The radiographic intensifying screen according to claim 7, wherein the organic silicon compound is an alkylalkoxysilane. 9. A radiographic intensifying screen which comprises a support, a fluorescent layer formed on the support, a resin layer provided by coating a resin solution on the fluorescent layer, and a protective layer formed by coating a solution containing a protective layer-forming resin on the resin layer, and wherein the resin layer contains a water repellent. 10. The radiographic intensifying screen according to claim 9, wherein the fluorescent layer contains a binder in an amount of 1 to 10 parts by weight per 100 parts by weight of a phosphor. 11. The radiographic intensifying screen according to claim 9, wherein the resin layer is thinner than the protective layer and has a thickness of from 0.3 .mu.m to 2.0 .mu.m. 12. The radiographic intensifying screen according to claim 9, wherein the water repellent is an organic silicon compound. 13. The radiographic intensifying screen according to claim 12, wherein the organic silicon compound is an alkylalkoxysilane. 14. A process for preparing a radiographic intensifying screen, which comprises coating a solution containing a water repellent on a fluorescent layer previously formed, drying, further coating a solution containing a protective layer-forming resin thereon, and drying. 15. A process for preparing a radiographic intensifying screen, which comprises coating a phosphor solution containing a water repellent on a support to form a fluorescent layer thereon, drying, further coating a solution containing a protective layer-forming resin thereon, and drying, wherein the water repellent is distributed so that the proportion of the water repellent in the fluorescent layer near the interface to the protective layer is higher than the proportion of the water repellent in the fluorescent layer near the interface to the support. 16. A process for preparing a radiographic intensifying screen, which comprises coating a resin solution containing a resin, a water repellent and an organic solvent on a fluorescent layer previously formed, drying to form a resin layer, further coating a solution containing a protective layer-forming resin thereon, and drying to form a protective layer. 17. The process according to claim 16, wherein the resin solution and the solution containing the protective layer-forming resin are respectively coated in such amounts as to make the resin layer thinner than the protective layer and to make the thickness of the resin layer from 0.3 .mu.m to 2.0 .mu.m. 18. The process according to claim 16, wherein the resin of the resin layer is not dissolved in the solution containing the protective layer-forming resin. 19. The process according to claim 16, wherein the viscosity of the resin solution is higher than the viscosity of the solution containing the protective layer-forming resin. |
058752218 | summary | CROSS-REFERENCE TO RELATED APPLICATION This application is a continuation of copending international application PCT/DE96/00014, filed Jan. 8, 1996. BACKGROUND OF THE INVENTION Field of the Invention The invention relates to a method of operating a boiling-water reactor which is an unstable state as a result of local oscillation of a physical variable (in particular of the power or of the neutron flux associated therewith). In addition, the invention relates to a device for carrying out this method and to a method and a device for monitoring the unstable reactor state. The nuclear fission determining the power of a nuclear reactor is controlled by moving absorber elements into the reactor core in order to attenuate the neutron flux. In this arrangement, measuring lances having sensors for the flux of thermal neutrons are distributed over the reactor core, in order to register the current state. In order to adjust a desired operating state, it is also necessary for the throughput of coolant (cooling water), which serves at the same time as a moderator, to be adapted to the respective state. The coolant enters in liquid phase into the reactor core from below, flows through the fuel elements, in which it partially evaporates, and emerges from the core as a vapor phase/liquid phase mixture, as a result of which the fuel/moderator ratio in the various parts of the fuel elements is changed. At the same time, however, the flow conditions are changed, in particular the location at which the single-phase flow, with which the liquid coolant enters the fuel elements, changes into the two-phase flow of the liquid/vapor mixture. In this case, at high power and low coolant throughput, unstable conditions have been observed in which this phase boundary goes into an oscillating motion, which results in a pulsation of the moderator density and the power, which has a bearing on the cooling capacity and the movement of the phase boundary. In this case, periodic temperature fluctuations with considerable peak values may occur in the fuel elements. The permissible power maximum of the fuel elements is mainly limited by the temperature resistance of the materials used in the fuel elements. If an upper temperature limit is exceeded, the materials loose their mechanical, chemical and physical properties and can undergo irreversible changes, which can force an exchange of the fuel elements. Therefore, care must be taken that this thermal-hydraulic upper power threshold (and hence a thermal-hydraulic threshold value A.sub.th of the neutron flux) in the reactor is not exceeded. Safety provisions in the reactor operation therefore call for a rapid shutdown of the reactor (so-called "SCRAM"), in the event the threshold value is exceeded. In such an emergency program, all the control rods are rapidly moved in and the corresponding cooling capacity is set. Following such a SCRAM, the reactor is restarted according to a predetermined startup program, so that there is a considerable disturbance to the reactor operation. In addition, the fuel elements have to be changed for safety reasons, if the thermal-hydraulic threshold value has been reached many times or over a relatively long period of time. The art is therefore concerned with detecting and damping an unstable state of this type as early as possible, before the power pulsations reach the vicinity of the thermal-hydraulic threshold value. It has been shown that these pulsations always occur in a frequency range between about 0.3 and 0.7 Hz and have a very constant frequency. The method described in U.S. Pat. No. 5,174,946 to Watford et al. (=EP 0 496 551) for monitoring the power fluctuation band for nuclear reactors is based on that fact. That process utilizes the flux as a measured variable for the unstable state caused by the local oscillation of a physical variable, the measuring lances mentioned ("local power range monitor-strings", LPRM strings) being used for this flux measurement. Each such lance normally contains four sensors, whose signals are observed anyway for power control purposes, then further processed and documented. Each of these four sensors in each measuring lance is used, two sensors being assigned to a first monitoring system, the two remaining sensors being assigned to a redundant second monitoring system. Each monitoring system thereby contains two monitoring channels, each sensor signal of a measuring lance being assigned to a different monitoring channel. Different subdivisions of the reactor into individual regions ("monitoring cells") are in this case based on the two monitoring channels of a system, each cell being bounded by four measuring lances in order to form a corresponding region signal. Depending on the location of the measuring lance in the core (in the interior of the core or at the edge of the core), a sensor signal in each monitoring channel belongs to two, three or four cells. As a result of this multiple use of the sensor signals, it is intended to achieve the situation where virtually the state of each individual fuel element can be monitored and identified by means of the influence which it has on the sensor signals of the individual cells. To this end, provision is made that an alarm is set in a system only when both monitoring channels respond. Although it is sufficient for the alarm to be given by one of the two systems, only simple redundancy is provided thereby. A further disadvantage is that virtually all the monitoring channels are affected by an erroneous measurement or a complete failure of a measuring lance, it being possible in the case of an edge position of the measuring lance, for example, that simultaneously a plurality of cells are no longer being monitored properly. The state of the individual cells (regions) is monitored by initially monitoring in a plausibility control whether the individual sensor signal exceeds a specific lower threshold value and is operating properly. In the case of a sensor defect, the signals belonging to this cell are not evaluated further. By means of summing all the sensor signals of a region, a current region signal is formed which is suppressed, however, if (for example as a result of an erroneous measurement) a plausibility monitoring yields the fact that the region signal does not achieve a predefined minimum value. The region signal is then filtered and related to an average over time, the time constant of which is greater than a period of the oscillation, so that a relative current region signal is produced which indicates by how many percent the current power of the region lies above or below the average. If this current value exceeds a power limit (for example 120%), a check is then made as to whether this is a once-off transition state (so-called "transient") which for example constitutes only an aperiodic transition to a new operating state predefined by the control, without exciting an oscillation. In this case, this is not therefore a critical oscillation in the frequency band from 0.3 to 0.7 Hz, so that no intervention is carried out as long as a threshold value A.sub.max, lying in the vicinity of the thermal-hydraulic threshold value A.sub.th, is not reached. In order to detect the critical oscillation, instead an examination is made to see whether, in a time interval corresponding to this critical frequency band, the value does not also fall below a corresponding threshold value (e.g. 80%) following the exceeding of a limiting value A.sub.o as is necessary for an oscillation. If it is determined in this way that--corresponding to an oscillation--a lower extreme value follows an upper extreme value of the flux, a check is further made as to whether another upper extreme value follows this lower extreme value, and whether this following upper extreme value exceeds an alarm value which lies above the extreme value detected first by a predefined factor (e.g. 1.3). If this is so, then after this one oscillation period it is already concluded that there is a growing, i.e., increasing oscillation, in which the exceeding of A.sub.th is threatened, and the SCRAM is initiated even before the value A.sub.max is reached. With an eye to the present invention, reference is made at this point that, although the above-described prior art monitors whether the oscillation is growing at a rate lying above the predefined factor (here 1.3), the growth (rate of increase) of the extreme values is not itself measured. This factor (1.3) is also relative in as much as it is related to the extreme value detected first, but it independent of the rate of increase. In addition, reference is made to the fact that although it is checked whether the time interval between the detected extreme values corresponds to the critical frequency band of 0.3 and 0.7 Hz, no check is made as to whether the next extreme value A.sub.n+1 follows in practice at the same interval DT.sub.n, which is given by the previously detected upper extreme value (denoted A.sub.n-1, point in time T.sub.n-1) and the presently detected lower extreme value (A.sub.n, point in time T.sub.n), after this point in time T.sub.n. Those skilled in the art of reactor control and monitoring will appreciate that the usual techniques for the monitoring and documentation of the sensor signals apply and they will therefore readily be able not only to register the extreme values A.sub.n-1, A.sub.n, A.sub.n+1 . . . but also the points in time T.sub.n-1, T.sub.n, T.sub.n+1 . . . at which these extreme values occur. The person in charge of monitoring could therefore readily suppress the corresponding region signal if the time interval DT.sub.n =T.sub.n -T.sub.n-1 deviates significantly (for example 0.1 seconds) from the time interval DT.sub.n+1 =T.sub.n+1 -T.sub.n. However, U.S. Pat. No. 5,174,946 contains no advice on this point. In the state of that prior art, therefore, no attention is initially paid to an oscillation whose (unmeasured) rate of increase lies below the set factor (1.3); rather, intervention is considered in the reactor operation only when its extreme values exceed the threshold value A.sub.max. Only rapidly increasing oscillations cause this extremely critical state to be recognized in good time and to the initiation of suitable countermeasures. Apparently, it is assumed that slowly increasing oscillations inherently decay by themselves and normally do not require a SCRAM. To be specific, that prior art provides as counter-measure only to damp the oscillation by means of rapidly moving in virtually all the control rods (total SCRAM). That is to say, apart from the SCRAM, this strategy provides no further measure for damping the oscillation and does not reduce the probability of the SCRAM either, which constitutes a considerable intervention in the reactor operation. Instead, in the event that there is a rapidly increasing oscillation, damping only takes place earlier (i.e., below A.sub.max). As a result, only the thermal loading of the fuel elements is reduced. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a method and device for operating a reactor in an unstable state, which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which improves the oscillation detection and damping so as to entirely obviate any SCRAM, i.e., to manage without an intervention in the reactor operation, or with an intervention which is the least disturbing. It is a further object to allow the monitoring of the critical state with a system and method which is least susceptible to interference. With the foregoing and other objects in view there is provided, in accordance with the invention, in a reactor operated in accordance with operationally dependent input parameters, a method of operating the reactor which is unstable as a result of an oscillation of an internal physical variable, which comprises: measuring the physical variable during at least two oscillations and calculating at least one measured value for a rate of increase of the oscillation; and PA1 deciding, in dependence on the measured value, whether a stabilization strategy is to be initiated with changed input parameters for damping the instability or a reactor operation is to be continued with unchanged input parameters. PA1 a system selection stage, a plurality of region selection stages connected to the system selection stage, a given number of region monitoring stages connected to each the region selection stage, and a sensor stage connected to each the region monitoring stage with a plurality of sensors strategically disposed in regions of a reactor core of a boiling water reactor, wherein PA1 a) measured signals supplied by the sensors to a respective the region monitoring stage are combined into a region signal for the physical variable; each the region signal is monitored in the respective the region monitoring stage in accordance with a monitoring criterion, and a region signal containing a region monitoring signal is output by each region monitoring stage; PA1 b) each region signal is connected to at least one of the region selection stages, and the region selection stages forming respective system monitoring signals from a predefined minimum number of region monitoring signals; and PA1 c) the region selections stages each outputting a respective system monitoring signal to the system selection stage, and the system selection stage outputting an output monitoring signal according to a predefined minimum number of systems. PA1 a) a plurality of sensors disposed in a plurality of regions of a reactor core of a boiling-water reactor, the sensors measuring a physical variable of the reactor core and outputting output signals, the output signals of a plurality of the sensors of a given region being combined into an associated region signal; PA1 b) a plurality of evaluation stages each receiving a respective region signal, the evaluation stages identifying in the region signal an occurrence of extreme values of the physical variable and, given an oscillation of constant frequency, determining a rate of increase of the extreme values in the respective region; and PA1 c) at least one monitoring stage receiving output signals from the evaluation stages, the monitoring stage setting an alarm signal when the extreme values of a predefined number of regions satisfy a local monitoring criterion which depends on the rate of increase of the extreme values By measuring the physical variable (i.e., the neutron flux, in the case of the thermally-hydraulically induced oscillations) the invention provides for the formation of local measured values in a plurality of regions of the reactor core, the measured values being assigned to the respective regions. The monitoring of the measured values leads to the formation of a current alarm stage from among a hierarchy of alarm stages with associated monitoring criteria, and the selection of the highest alarm stage, whose monitoring criterion is satisfied by the measured values in a predefined minimum number of the regions. (The monitoring criterion can in this case be composed of a plurality of individual conditions, for example the exceeding of separate threshold values for the amplitude and for the rate of increase of the extreme values.) Depending on the current alarm stage, a stabilization strategy is then initiated. As a stabilization strategy which belongs to a low-ranking alarm stage, provision is made to intervene in the operational control and regulation of the reactor only so as to block a removal of the control rods, as is envisaged in the case of an operational increasing of the reactor power: the power of the reactor cannot then be raised by the operating personnel of the reactor; instead only such control commands which correspond to the control of the reactor to a constant or decreasing power become effective in the reactor control system. In at least one higher-ranking alarm stage, provision is made as stabilization strategy for a plurality of control rods to be introduced into the core in the sense of a reduction in the reactor power (alarm stage I). Advantageously, at least two higher-ranking alarm stages (alarm stage II and alarm stage III) are provided, in alarm stage II only a plurality of control rods, corresponding to a fraction of the total number, being moved into the core slowly and in such a way as corresponds to an operational reduction in the power (that is to say the reactor control system performs an operational reduction in the power, even if, for example, a higher power consumption would intrinsically require a higher reactor power and the operating personnel wish to increase the reactor power). In the second higher-ranking alarm stage (alarm stage III)--in a manner similar to the case of a total rapid shutdown of the reactor (total SCRAM)--control rods are moved in rapidly, however likewise not all thereof but only some of the control rods being involved ("partial SCRAM"). A total SCRAM is then no longer necessary, but an option for an alarm stage IV which triggers the SCRAM can be retained. In particular, during the monitoring of the measured values, at least two periods of the oscillation are evaluated, so that the reactor is therefore initially further operated in an unchanged manner, although an oscillation is already indicated. Furthermore, a method of operating a reactor which is unstable as a result of oscillation of a physical variable occurring in the core makes provision, by measuring the physical variable, for forming a measured value which registers the rate of increase of the oscillation (if appropriate, also further measured values). Depending on this measured value, a decision is made as to whether a stabilization strategy should be initiated in order to damp the instability or the reactor is initially further operated in accordance with measured values entered as a function of operation. In particular, in this case the reactor can continue to be operated for at least two more oscillations during the measurement of the rate of increase, without an intervention being made in the reactor control system--provided that no measured value reaches a threshold value which calls for the initiation of a total SCRAM. Thus, for example, it is possible that when a threshold value A.sub.max for the oscillation amplitudes is exceeded, the SCRAM--corresponding to the highest alarm stage IV--is initiated only at high rates of increase, but at low rates of increase the reactor is still operated with relatively high amplitudes, since in the case of amplitudes which are growing so weakly, a SCRAM which is initiated only later (in the event that the oscillation then does not intrinsically decay) still has sufficient time to become effective before A.sub.th is reached. A threshold value, dependent on the rate of increase, is preferably predefined for the extreme values of the oscillating physical variable, and the stabilization strategy is triggered if the extreme values exceed this threshold value. However, a threshold value for the rate of increase can also be predefined, the stabilization strategy then being initiated when the rate of increase exceeds this threshold value. In a similar embodiment of the invention, a number of oscillations can be predefined, the number depending on the rate of increase, and the stabilization strategy can then be triggered only when the oscillation of the physical variable persists over the duration of these oscillation periods. In accordance with an added feature of the invention, a plurality of stabilization strategies are provided, from which the stabilization strategy to be triggered is selected as a function of the rate of increase. The (unstable) state of the reactor core is monitored with a plurality of sensors which are strategically distributed about the core. The sensor locations are divided into a plurality of regions of the reactor core and the sensors measure the behavior of the physical variable in those regions. The output signals of the sensors are combined into a number Mp of region channels and each region channel is assigned a region and sensors arranged therein for generating a region signal. The region signals are then combined into a number P of system channels, with a plurality of region channels being assigned to a system channel, in that they generate a system signal. The system signals are finally assigned to an output channel and they generate an output signal. By means of monitoring stages and selection stages, in this case an alarm output signal is set in the output signal as soon as, at least in a predefined number N.sub.p of the system channels, particularly in a minimum number N.sub.mp of region channels of the system, a monitoring criterion is satisfied over a plurality of oscillation periods. In this case, the output signal of each sensor influences a maximum of one single region signal and each region signal influences a maximum of one system signal. The region signals of a system channel are in each case formed from the output signals of sensors which are located in regions which are distributed over the cross section of the reactor core in such a way that the regions which are adjacent to such a region contain sensors whose output signals are assigned to region channels of other system channels. The invention thus effectively dispenses with multiple evaluations and region overlaps. Although each individual fuel element is no longer as precisely monitored as in the Watford et al. patent, experience and model calculations with unstable states have shown that it is always relatively large parts of the reactor, but not isolated fuel elements, which begin oscillating. In other words, fine resolution of the measured value registration is not necessary. In addition, the redundancy and interference immunity of the registration is increased. With the above and other objects in view, there is further provided, in accordance with the invention, a device for monitoring a reactor core of a boiling-water reactor with regard to local oscillations of a physical variable causing an unstable state of the reactor. The device comprises: There is further provided, in accordance with the invention, a device for monitoring a reactor core of a boiling-water reactor with regard to a state which is unstable as a result of local oscillation of a physical variable in the reactor core, comprising: In other words, there is provided a system selection stage, a number P of region selection stages, for each region selection stage a number Mp of region monitoring stages and for each region monitoring stage a sensor stage having a plurality of sensors which are arranged inside a region of the core and are assigned to this region monitoring stage. The device is constructed in such a way that the sensors which are respectively assigned to a region monitoring stage supply measured signals for the physical variable which are combined into a region signal, and each region signal is monitored in accordance with a monitoring criterion in the region monitoring stage assigned to the sensors. Each region monitoring stage supplies a region signal which contains a region monitoring signal. Each region monitoring signal is connected to at least one region selection stage which forms a system monitoring signal from a predefined minimum number of region monitoring signals. Each system monitoring signal is then fed to the system selection stage; the latter supplies an output monitoring signal by means of a predefined minimum number of system monitoring stages. The system sensors which are strategically distributed about a plurality of regions of the reactor core for measuring the physical variable. The output signals of a plurality of sensors of a region are combined into an associated region signal. Each region signal is assigned an evaluation stage, which identifies in the region signal the occurrence of extreme values of the physical variable (in particular over a plurality of oscillation periods) and, given an oscillation of constant frequency and appropriate duration, determines the rate of increase of the extreme values in this region. The evaluation stages are assigned at least one monitoring stage which sets an alarm signal as soon as the extreme values at least in a predefined number of regions satisfy a local monitoring criterion which depends on the determined rate of increase. With a view to the proposed stabilization criteria, a device for monitoring the local oscillations can contain sensors for measuring the physical variable, which sensors are arranged in a plurality of regions of the reactor core, and the output signals of a plurality of sensors of a region being combined into an associated region signal. Each region signal is then assigned an evaluation stage which identifies the occurrence of an oscillation of constant frequency in the region signal. The evaluation stages are assigned an output monitoring stage which selects an alarm stage from a hierarchy of alarm stages in accordance with predefined monitoring criteria for the oscillations identified in at least a predefined number of region signals. In this case, the output monitoring stage, corresponding to the selected alarm stage, defines a point in time (or at least the criteria for the point in time) at which an emergency instruction is output to initiate a stabilization strategy corresponding to the alarm stage. This point in time can be predefined, for example, by means of a number of oscillation periods which are allowed to elapse before the initiation of a stabilization measure. However, by this means it can also be defined that, depending on the instantaneous current values (for example current values of the rate of increase) a threshold value (for example a threshold value for the amplitude) is defined, which leads to the triggering of the stabilization measure at a later point in time, at which a monitored current value (e.g., the amplitude) then exceeds this predetermined threshold value. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a method and device for operating a reactor in an unstable state, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
abstract | An optical system is disclosed that includes a plurality of elements arranged to image radiation at a wavelength λ from an object field in an object surface to an image field in an image surface. The elements include mirror elements have a reflective surface formed by a reflective coating positioned at a path of radiation. At least one of the mirror elements has a rotationally asymmetrical reflective surface deviating from a best-fit rotationally symmetric reflective surface by about λ or more at one or more locations. The elements include an apodization correction element effective to correct a spatial intensity distribution in an exit pupil of the optical system relative to the optical system without the apodization correcting element. The apodization correction element can be effective to increase symmetry of the spatial intensity distribution in the exit pupil relative to the optical system without the apodization correcting element. |
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050420593 | summary | BACKGROUND OF THE INVENTION 1. Field of The Invention This invention relates to optical elements for radiation which are particularly useful for X-ray spectroscopy, neutron spectroscopy and the like spectroscopic analyses and which are made of artificial graphite. 2. Description of The Prior Art As is known in the art, optical elements which are usually used in optical instruments for X-rays such as X-ray spectroscopes, X-ray microscopes and the like make use of the Bragg reflection of crystals although, in a specific case, total reflection of X-rays passing very closely to a reflection surface may be utilized. The crystals used for the above purpose should have a complete crystal structure and should be obtained in a desired size. Moreover, such crystals should have a small absorption coefficient for X-rays and should have appropriate flexibility when applied to curved spectroscopes. One type of crystal which satisfies the above requirements is graphite. Graphite has a small absorption coefficient against X-rays and has been often employed as an optical element for X-rays. However, a single crystal of natural graphite with a large area cannot be obtained. Accordingly, it is usual to artificially obtain graphite crystals by hot processing of a pyrolytic deposit of hydrocarbon. For instance, there is known as artificial graphite compression annealed pyrographite (CAPC) or high-oriented pyrographite (HOPG) commercially sold from Union Carbide Inc. These graphite products are produced by pyrolyzing a gaseous hydrocarbon at a temperature of about 1000.degree. C. to obtain graphite crystals and annealing the crystals at 3600.degree. C. over a long time of, say, several weeks under pressure. As is known, the Bragg equation is expressed as 2d sin .theta.=.lambda. where d is the distance between the successive lattice planes, .theta. is a reflection angle, and .lambda. is a wavelength of the reflected X-ray. It is stated that with the graphite of Union Carbide Inc., where a monochromatic X-ray, e.g. a K .alpha. line of Cu radiation (.lambda.=1.5418 .ANG.), reflects on the (002) plane, the distance between the lattice planes, d, is close to the distance between the single crystals of the graphite, i.e. d=3.354 .ANG.. The half-value width of the reflection line, .DELTA..delta..sub.002, is stated to be about 0.7.degree.. However, these artificial graphite products involve the problem that annealing under very high temperature and long time conditions at a high pressure is required as mentioned above, thus the production process being complicated with high production costs. For focussing X-rays, it is usual to appropriately bend a single crystal plate of silicon or to form a curved lens of graphite by machining. These are also complicated in the fabrication with high production costs. For the purpose of providing artificial graphite sheets with a large area which are simply fabricated without resorting to complicated procedures such as compression annealing and are thus inexpensive but which have complete crystallinity and good flexibility, we proposed in U.S. Pat. No. 4,788,703 (corresponding European Patent Laid-open Application No. 219,345) graphitization of a polyphenylene oxadiazole (POD) by treatment at 2800.degree. C. or higher. The graphitized sheet was flexible and was found to be suitable as a radiation optical element such as for X-rays. The graphitized POD obtained by treating starting POD at a normal pressure at temperatures not lower than 2800.degree. C. has the following physical properties. (1) Reflection lines against CuK .alpha. (1.5418 .ANG.) are those corresponding only to (002), (004) and (006) planes. (2) The reflection angle (2.theta.) at the (002) plane is 26,576.degree. and the distance between the lattice planes, d, is 3,354 .ANG., which were in coincidence with those of the single crystal of graphite. (3) The half-value widths of the reflection line (having a center at 2.theta.=26.576.degree.) at the (002) plane were, respectively, 2.0.degree. and 0.14.degree. for thermal treatment at temperatures of 2800.degree. C. and 3000.degree. C. (4) The graphitized POD had flexibility and the area or size of the product could be increased as desired depending upon the area or size of a starting POD sheet and the size of a thermal treatment furnace. The radiation optical element using the graphitized POD sheet exhibit good characteristics when applied as an X-ray lens, an X-ray monochromater or an optical element for neutron spectroscopy. However, the graphitized POD has the problem that a rocking characteristic which is the most important characteristic when it is applied as a radiation optical element is not satisfactory. For instance, the rocking characteristic of a graphitized POD product treated at 3000.degree. C. is 6.9.degree., which is unsatisfactory for use as a radiation optical element although such a characteristic may be further improved by hot pressing. Moreover, the graphitized POD is also unsatisfactory with respect to the reflectivity of radiation. The radiation reflectivity has the relation with the crystallinity along the c axis of graphite. It is known that if the crystallinity along the c axis is too good, an X-ray once passing into the crystal will suffer internal reflection, causing a total reflectivity to be lowered. In order to realize a good reflection efficiency, the crystallinity along the c axis of graphite should not be too good or too bad but is required to have an appropriate value. The graphitized POD is good in crystallinity along the c axis, which adversely influences the reflection efficiency on radiation. In recent years, X-rays or so-called soft X-rays having a wavelength of from approximately 9 or 10 .ANG.to several hundreds angstroms are being utilized in the field of lithography of semiconductor, on which industrial importance is placed. Among soft X-ray elements dealing with an X-ray having about 10 angstroms to several hundreds angstroms, those elements using diffraction should have a lattice distance, contributing to the diffraction, which corresponds to an intended wavelength, say approximately 10 angstroms to several hundreds angstroms. With graphite currently used as the X-ray element, the lattice distance contributing to the diffraction is 3.354 .ANG.. Accordingly, such graphite cannot be used for the purpose of reflection of soft X-rays. SUMMARY OF THE INVENTION It is therefore an object of the invention to provide an optical element for radiation which can solve the drawbacks of the prior art elements and can be simply manufactured with low production costs. It is another object of the invention to provide a radiation optical element which has better rocking and reflectivity characteristics than known elements using graphite from polyphenylene oxadiazole set forth before. It is a further object of the invention to provide an optical element for soft X-rays whose wavelength is in the vicinity of 10 angstroms as well as hard X-rays. It is a still further object of the invention to provide a radiation optical element which comprises a graphite sheet having a desired thickness capable of dealing with not only X-rays, but also neutron rays. It is another object of the invention to provide a radiation optical element made of a graphitized product of a polymer in the form of a sheet which may be used in combination with a suitable substrate. The optical element for radiation of the present invention is adapted for use in a radiation optical system which includes a radiation source capable of emitting a radiation and a means for receiving the radiation generated from the radiation source through the optical element. The optical element may be used as a convergent lens, a monochromator, an analyzer, a filter or the like means for radiations. The radiations may include X-rays including so-called soft and hard X-rays, neutron rays and the like. In accordance with one embodiment of the invention, there is provided a radiation optical element which comprises a graphite film or sheet obtained by thermally treating or pyrolyzing a film of a polymer in vacuum or in an inert gas at a temperature of not lower than 2800.degree. C. at a pressure of not lower than 4 kg/cm.sup.2. The polymer should be a member selected from polyphenylene oxadiazole, polybenzothiazole, polybenzobisthiazole, polybenzooxazole, polybenzobisoxazole, polypyromellitimide, polyphenylene-isophthalamide, polyphenylene-benzoimidazole, polyphenylene-benzobisimidazole, polythiazole and poly-p-phenylene-vinylene. The resultant graphite film is significantly improved in rocking characteristic and reflection efficiency. The polyphenylene oxadiazole is included as the starting polymer. As will become apparent from examples, the graphite film obtained from the polyphenylene oxadiazole (POD) is better in the characteristic and efficiency than a known graphite film of POD obtained by thermal treatment at temperatures not lower than 2800.degree. C. at a normal pressure. The process of pyrolysis of POD under a certain pressure has not been known as providing a final graphite product of better characteristic properties. If a graphite sheet or block having a desired thickness is necessary, such a sheet can be obtained by pressing a plurality of polymer films or a plurality of granite films at a final temperature of not lower than 2800.degree. C. for a sufficient time. In accordance with another embodiment of the invention, there is also provided a radiation optical element which comprises a graphite film interacted or intercalated with a metal halide, thereby forming a film of a metal halide-intercalated graphite compound. The graphite film is obtained from a polymer which is thermally treated in the same manner as in the first embodiment. The polymer used is a member selected from those defined in the first embodiment. When a plurality of the intercalated graphite films are pressed, a block or sheet of a desired thickness can be readily obtained because of the good adhesiveness of the intercalated graphite films. This intercalated film is flexible and can be conveniently provided on or along a curved surface. In accordance with a further embodiment of the invention, there is also provided a radiation optical element which comprises at least one graphite film obtained in the first embodiment and at least one metal halide intercalated graphite film which are superposed and press bonded. In this embodiment, it is possible and even usual to alternately superpose a number of graphite films and a number of the intercalated graphite films or to sandwich a number of the intercalated graphite films between two graphite films with or without further superposition of a plurality of the sandwiched sheets thereby forming a block of a desired thickness in the order of millimeters. In all the embodiments, the film sheets or blocks may be used as they are. If a thin graphite or intercalated graphite compound film is used, it is favorable to form it on a substrate so as to impart mechanical strength thereto. |
description | The present application is related to and claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Related Applications”) (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Related Application(s)). All subject matter of the Related Applications and of any and all parent, grandparent, great-grandparent, etc. applications of the Related Applications is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of U.S. patent application Ser. No. 12/590,447, entitled SYSTEMS AND METHODS FOR CONTROLLING REACTIVITY IN A NUCLEAR FISSION REACTOR, naming Charles E. Ahlfeld, Ehud Greenspan, Roderick A. Hyde, Nathan P. Myhrvold, Joshua C. Walter, Kevan D. Weaver, Thomas Allan Weaver, Lowell L. Wood, Jr., and George B. Zimmerman as inventors, filed Nov. 6, 2009 now U.S. Pat. No. 9,190,177, The United States Patent Office (USPTO) has published a notice to the effect that the USPTO's computer programs require that patent applicants reference both a serial number and indicate whether an application is a continuation or continuation-in-part. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette Mar. 18, 2003, available at http://www. uspto.gov/web/offices/com/sol/og/2003/week 11/patbene.htm. The present Applicant Entity (hereinafter “Applicant”) has provided above a specific reference to the application(s) from which priority is being claimed as recited by statute. Applicant understands that the statute is unambiguous in its specific reference language and does not require either a serial number or any characterization, such as “continuation” or “continuation-in-part,” for claiming priority to U.S. patent applications. Notwithstanding the foregoing, Applicant understands that the USPTO's computer programs have certain data entry requirements, and hence Applicant is designating the present application as a continuation-in-part of its parent applications as set forth above, but expressly points out that such designations are not to be construed in any way as any type of commentary and/or admission as to whether or not the present application contains any new matter in addition to the matter of its parent application(s). The present application is related to controlling reactivity in a nuclear fission reactor. Illustrative embodiments provide a reactivity control assembly for a nuclear fission reactor, a reactivity control system for a nuclear fission reactor having a fast neutron spectrum, a nuclear fission traveling wave reactor having a fast neutron spectrum, a method of controlling reactivity in a nuclear fission reactor having a fast neutron spectrum, methods of operating a nuclear fission traveling wave reactor having a fast neutron spectrum, a system for controlling reactivity in a nuclear fission reactor having a fast neutron spectrum, a method of determining an application of a controllably movable rod, a system for determining an application of a controllably movable rod, and a computer program product for determining an application of a controllably movable rod. The foregoing summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. The present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structures)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Illustrative Reactivity Control Assembly Referring now to FIG. 1A and given by way of overview, an illustrative reactivity control assembly 10 for a nuclear fission reactor (not shown) is shown. A reactivity control rod 12 includes neutron absorbing material 14 configured to absorb neutrons (not shown). At least a portion of the neutron absorbing material 14 includes fertile nuclear fission fuel material 16. At least one sensor 18 is physically associated with the reactivity control rod 12. The sensor 18 is configured to sense status of at least one reactivity parameter associated with the reactivity control rod 12. Illustrative details will be set forth below by way of non-limiting examples. It will be appreciated that the reactivity control rod 12 may be any type of suitable reactivity control rod. In some embodiments the reactivity control rod 12 may be a stand-alone reactivity control rod. That is, in such an arrangement the reactivity control rod 12 is not grouped into an assembly with other rods, such as nuclear fission fuel rods and/or other reactivity control rods. In some other embodiments, the reactivity control rod 12 may be part of an assembly that includes nuclear fission fuel rods and/or other reactivity control rods. It will also be appreciated that the reactivity control rod 12 may have any physical shape as desired for a particular application. Given by way of non-limiting examples, in various embodiments the reactivity control rod 12 may have a cross-sectional shape that is square, rectangular, circular, ovoid, or any shape as desired. In various embodiments the reactivity control rod 12 may be embodied as a blade, and may have any cross-sectional shape as desired, such as a rectangle, an “X”, a “+”, or any other shape. The reactivity control rod 12 may have any shape that is suited for the nuclear fission reactor in which the reactivity control rod 12 is to be used. No limitation regarding shape of the reactivity control rod 12 is implied, and none should be inferred. In some embodiments the neutron absorbing material 14 may be configured to absorb fast spectrum neutrons. For example, the neutron absorbing material 14 may have an absorption cross-section that permits absorption of fast spectrum neutrons—that is, neutrons having an energy level on the order of at least around 0.11 MeV. Given by way of non-limiting example, the neutron absorbing material 14 may have an absorption cross-section on the order of around 10 barns or less. In some embodiments the fertile nuclear fission fuel material 16 may serve as the component of the neutron absorbing material 14 that absorbs the fast neutrons. In some other embodiments, other component(s) of the neutron absorbing material 14 may also serve as additional component(s) of the neutron absorbing material 14 (in addition to the fertile nuclear fission fuel material 16) that absorbs the fast neutrons. Illustrative details regarding fertile nuclear fission fuel material 16 and other components of the neutron absorbing material 14 will be set forth below. In some applications, it may be desirable to maintain the neutron spectrum of a nuclear fission reactor within the fast neutron spectrum. Given by way of non-limiting examples, the reactivity control assembly 10 may be used to help control reactivity in a fast nuclear fission reactor, such as without limitation a traveling wave reactor or a fast breeder reactor, like a liquid metal fast breeder reactor or a gas-cooled fast breeder reactor, or the like. To that end, in some other embodiments the neutron absorbing material 14 may be configured to reduce moderation of neutrons. For example, the neutron absorbing material 14 may have a suitably large atomic mass that can help reduce the amount of slowing down of fast spectrum neutrons. As such, a reduction may be made in softening of the neutron spectrum from the fast neutron spectrum toward neutron spectrums having neutron energy levels less than around 0.1 MeV, such as an epi-thermal neutron spectrum or a thermal neutron spectrum. It will be appreciated that, given by way of non-limiting examples, elements of the actinide series, such as without limitation uranium and thorium, present a sufficiently large atomic mass to help reduce moderation of neutrons. In some embodiments the fast spectrum neutrons may be part of a nuclear fission traveling wave. A nuclear fission traveling wave may also be referred to as a nuclear fission deflagration wave. Non-limiting examples of initiation and propagation of a nuclear fission traveling wave is described in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; and U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the entire contents of which are hereby incorporated by reference. The fertile nuclear fission fuel material 16, that is included in the neutron absorbing material 14, can include any type of fertile nuclear fission fuel material as desired for a particular application. For example, in some embodiments the fertile nuclear fission fuel material 16 may include uranium, such as 238U. It will be appreciated that the absorption cross-spectrum for fast neutrons of 238U is on the order of around 0.2 barns. In some other embodiments the fertile nuclear fission fuel material 16 may include thorium, such as 232Th. It will be appreciated that the absorption cross-spectrum for fast neutrons of 232Th is on the order of around 0.2 barns. The fertile nuclear fission fuel material 16 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. In some applications it may be desirable to soften the neutron spectrum from the fast neutron spectrum toward neutron spectrums having neutron energy levels less than around 0.1 MeV, such as an epi-thermal neutron spectrum or a thermal neutron spectrum. For example, in such applications the reactivity control assembly 10 may be used to help control reactivity in a thermal nuclear fission reactor, such as without limitation a pressurized water reactor. As another example, in some other applications the reactivity control assembly 10 may be used to help control reactivity in a fast nuclear fission reactor in which it is desired to soften the neutron spectrum to reduce irradiation damage. To that end and referring now to FIGS. 1B-1G, in some embodiments the reactivity control rod 12 may also include neutron moderating material 20 in addition to the fertile nuclear fission fuel material 16. The neutron moderating material 20 may include any suitable neutron moderating material as desired for a particular application. Given by way of non-limiting example, the neutron moderating material 20 may include any one or more of hydrogen, deuterium, helium, lithium, boron, carbon, graphite, sodium, lead, and the like. When the neutron moderating material 20 is provided, the neutron moderating material 20 may be distributed within the reactivity control rod 12 in any manner as desired for a particular application. For example and as shown in FIGS. 1B-1F by way of illustration and not of limitation, in some embodiments the neutron moderating material 20 may be substantially heterogeneously distributed within the reactivity control rod 12. Given by way of non-limiting examples, the neutron moderating material 20 may be heterogeneously distributed in disks 21 (FIGS. 1B and 1C). The disks 21 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 12 (as shown in FIG. 1B) or substantially transverse to the axial orientation of the reactivity control rod 12 (as shown in FIG. 1C). Given by way of further non-limiting examples, the neutron moderating material 20 may be heterogeneously distributed toward ends of the reactivity control rod 12 (as shown in FIG. 1D) or toward a middle of the reactivity control rod 12 (as shown in FIG. 1E). Given by way of a further non-limiting example, the neutron moderating material 20 may be provided as a rod follower 23 (as shown in FIG. 1F). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 1G, the neutron moderating material 20 may be substantially homogeneously distributed within the reactivity control rod 12. Referring now to FIGS. 1H-1M, in some embodiments the neutron absorbing material 14 may also include neutron absorbing poison 22 in addition to the fertile nuclear fission fuel material 16. The neutron absorbing poison 22 may include any suitable neutron absorbing poison as desired. For example and given by way of non-limiting examples, the neutron absorbing poison 22 may include any one or more of silver, indium, cadmium, gadolinium, hafnium, lithium, 3He, fission products, protactinium, neptunium, boron, and the like. The neutron absorbing poison 22 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. When the neutron absorbing poison 22 is provided, the neutron absorbing poison 22 may be distributed within the reactivity control rod 12 in any manner as desired for a particular application. For example and as shown in FIGS. 1H-1L by way of illustration and not of limitation, in some embodiments the neutron absorbing poison 22 may be substantially heterogeneously distributed within the reactivity control rod 12. Given by way of non-limiting examples, the neutron absorbing poison 22 may be heterogeneously distributed in disks 25 (FIGS. 1H and 1I. The disks 25 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 12 (as shown in FIG. 1H) or substantially transverse to the axial orientation of the reactivity control rod 12 (as shown in FIG. 1I). Given by way of further non-limiting examples, the neutron absorbing poison 22 may be heterogeneously distributed toward ends of the reactivity control rod 12 (as shown in FIG. 1J) or toward a middle of the reactivity control rod 12 (as shown in FIG. 1K). Given by way of a further non-limiting example, the neutron absorbing poison 22 may be provided as a rod follower 27 (as shown in FIG. 1L). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 1M, the neutron absorbing poison 22 may be substantially homogeneously distributed within the reactivity control rod 12. In some embodiments and referring now to FIGS. 1H-1P, an effect on reactivity achievable by the fertile nuclear fission fuel material 16 may equalized toward an effect on reactivity achievable by portions of the neutron absorbing poison 22. For example, such equalization may be desirable to mitigate localized flux peaking. It will be appreciated that such equalization may be effected regardless of whether the fertile nuclear fission fuel material 16 is distributed heterogeneously or homogeneously and regardless of whether the neutron absorbing poison 22 is distributed heterogeneously (FIGS. 1H-1L and FIGS. 1O-1P) or homogeneously (FIG. 1M). In some other embodiments and still referring to FIGS. 1H-1P, reactivity effect of the fertile nuclear fission fuel material 16 and reactivity effect of the neutron absorbing poison 22 may be locally tailored as desired for a particular application. For example, in some embodiments and as shown generally in FIG. 1N the reactivity control rod 12 has a region 24 and a region 26. It will be appreciated that the regions 24 and 26 may be located anywhere within the reactivity control rod 12 as desired. No limitation is implied, and is not to be inferred, by virtue of appearance in the drawings which are provided for illustration purposes only. A concentration 28 of the neutron absorbing poison 22 is disposed in the region 24 and a concentration 30 of the neutron absorbing poison 22 is disposed in the region 26. A concentration 32 of the fertile nuclear fission fuel material 16 is disposed in the region 24 and a concentration 34 of the fertile nuclear fission fuel material 16 is disposed in the region 26. It will be appreciated that concentration may be determined per volume basis, per area basis, or per length basis, as desired. It will be appreciated that reactivity effects of the concentrations 28 and 30 of the neutron absorbing poison 22 and reactivity effects of the concentrations 32 and 34 of the fertile nuclear fission fuel material 16 may be tailored as desired for a particular application. For example, in some embodiments and as shown in FIGS. 1H-1P a reactivity effect of the concentration 30 of the neutron absorbing poison 22 may be substantially equalized with a reactivity effect of the concentration 32 of the fertile nuclear fission fuel material 16. In some other embodiments and as also shown in FIGS. 1H-1P a reactivity effect of the concentration 28 of the neutron absorbing poison 22 may be substantially equalized with a reactivity effect of the concentration 34 of the fertile nuclear fission fuel material 16. In some other embodiments and as shown in FIGS. 1H-1P a reactivity effect of the concentration 30 of the neutron absorbing poison 22 may be different from a reactivity effect of the concentration 32 of the fertile nuclear fission fuel material 16. In other embodiments a reactivity effect of the concentration 28 of the neutron absorbing poison 22 may be different from a reactivity effect of the concentration 34 of the fertile nuclear fission fuel material 16. Other reactivity effects may be effected as desired. For example and as shown in FIGS. 1H-1P, in some embodiments a sum of reactivity effects of the concentration 28 of the neutron absorbing poison 22 and the concentration 32 of the fertile nuclear fission fuel material 16 may be substantially equalized toward a sum of reactivity effects of the concentration 30 of the neutron absorbing poison 22 and the concentration 34 of the fertile nuclear fission fuel material 16. In some other embodiments, reactivity effect is substantially constant between the region 24 and the region 26. If desired, concentration of the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may vary. For example and as shown in FIGS. 1O and 1P, in some embodiments concentration the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may change along a continuous gradient. Given by way of non-limiting example, as shown in FIG. 1O the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be provided in wedges 36 and 38, respectively, that abut each other along their hypotenuse 40. Given by way of another non-limiting example, as shown in FIG. 1P the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be provided in mated frustoconical sections 42 and 44, respectively. It will be appreciated that the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be provided in other suitable arrangements in which their concentrations change along a continuous gradient, and arrangements are not to be limited to those shown in FIGS. 1G and 1H by way of illustration and not of limitation. In some other embodiments, concentration of the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may change along a non-continuous gradient. For example, concentration of the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may change along a non-continuous gradient as a result of heterogeneous distribution as shown in FIGS. 1H-1L. In such cases, concentration of the neutron absorbing poison 22 can vary along a non-contiguous gradient because the neutron absorbing poison 22 is provided in discrete locations (as opposed to homogeneous distribution). Also in such cases, concentration of the fertile nuclear fission fuel material 16 can vary along a non-contiguous gradient because the fertile nuclear fission fuel material 16 is provided in discrete locations that are separated from each other by the discrete locations of the neutron absorbing poison 22. In some embodiments the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be spatially fixed relative to each other. That is, in such arrangements the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 do not physically move in relation to each other. However, in some other embodiments the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be spatially movable relative to each other. Given by way of non-limiting example and referring briefly to FIGS. 1H-1L and 1O-1P, any one or more of the discrete locations of the neutron absorbing poison 22, such as without limitation the disks 25, may be slidably received in the reactivity control rod 12 and may be moved in and out of the reactivity control rod 12 as desired by a suitable mechanism, such as a control rod drive mechanism (not shown) or the like. The sensor 18 may be physically associated with the reactivity control rod 12 in any suitable physical association as desired. For example, referring now to FIGS. 1A-1P and also to FIG. 1Q, in some embodiments physical association may include the sensor 18 being located within an interior 46 of the reactivity control rod 12. For example, the sensor 18 may be located via any suitable attachment method on an interior surface 48 of a cladding wall 50 of the reactivity control rod 12. As a further example and referring now to FIGS. 1A-1P and also to FIG. 1R, in some other embodiments physical association may include the sensor 18 being located proximate an exterior 52 of the reactivity control rod 12. For example, the sensor 18 may be located via any suitable method on an exterior surface 54 of the cladding wall 50. Any one or more of various reactivity parameters associated with the reactivity control rod 12 may be sensed with the sensor 18. Given by way of non-limiting examples, the sensed reactivity parameter may include any one or more of parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and/or neutron spectrum. The sensor 18 may include any suitable sensor that is configured to sense the reactivity parameter that is desired to be sensed. Given by way of non-limiting example, in some embodiments the sensor 18 may include at least one fission detector, such as without limitation a micro-pocket fission detector. In some other embodiments the sensor 18 may include a neutron flux monitor, such as without limitation a fission chamber and/or an ion chamber. In some embodiments the sensor 18 may include a neutron fluence sensor, such as without limitation an integrating diamond sensor. In some embodiments the sensor 18 may include a fission product detector, such as without limitation a gas detector, a β detector, and/or a γ detector. In some embodiments, when provided, the fission product detector may configured to measure a ratio of isotope types in fission product gas. In some embodiments the sensor 18 may include a temperature sensor. In some other embodiments the sensor 18 may include a pressure sensor. In some embodiments the sensor 18 may include a power sensor, such as without limitation a power range nuclear instrument. In some embodiments, if desired the sensor 18 may be replaceable. In some applications it may be desirable to mitigate effects of internal pressure within the reactivity control rod 12 exerted by fission products, such as fission product gases. In such cases and referring now to FIG. 1S, in some embodiments the reactivity control rod 12 may define at least one chamber 56 configured to accumulate fission products. For example, when provided the chamber 56 may include a plenum 58. In some embodiments the plenum 58 may be located at least one mean free path λT for fission-inducing neutrons from the fertile nuclear fission fuel material 16. In some embodiments a backflow prevention device 60, such as a check valve like a ball check vale or the like, may be provided to help prevent re-entry into the reactivity control rod 12 of fission product gases that have outgassed from the reactivity control rod 12. Referring now to FIG. 1T, in some embodiments a calibration device 62 configured to calibrate the sensor 18 may be provided. It will be appreciated that, when provided, the calibration device 62 suitably is a source having known characteristics or attributes of the reactivity parameter, discussed above, that is sensed by the sensor 18. Referring now to FIG. 1U, in some embodiments at least one communications device 64 may be operatively coupled to the sensor 18 as generally indicated at 66. The communications device 18 suitably is any acceptable device that can operatively couple the sensor 18 in signal communication with a suitable communications receiving device (not shown) as generally indicated at 68. Given by way of non-limiting examples, the communications device 64 may include an electrical cable, a fiber optic cable, a telemetry transmitter, a radiofrequency transmitter, an optical transmitter, or the like. Illustrative Reactivity Control System Referring now to FIG. 2A, an illustrative reactivity control system 210 is provided for a nuclear fission reactor (not shown) having a fast neutron spectrum. Given by way of overview, the reactivity control system 210 includes a reactivity control rod 212. The reactivity control rod 212 includes neutron absorbing material 214 configured to absorb fast spectrum neutrons. At least a portion of the neutron absorbing material 214 includes fertile nuclear fission fuel material 216. An actuator 217 is responsive to at least one reactivity parameter and is operationally coupled, as indicated generally at 219, to the reactivity control rod 212. Illustrative details will be set forth below by way of non-limiting examples. The actuator 217 may be responsive to any one or more of various reactivity parameters as desired for a particular application. In some embodiments, the reactivity parameter may include any one or more reactivity parameter of the nuclear fission reactor. In some other embodiments the reactivity parameter may include any one or more reactivity parameter of the reactivity control rod 212. Given by way of non-limiting examples, the reactivity parameter may include any one or more of parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and neutron spectrum. As mentioned above, the nuclear fission reactor (not shown) has a fast neutron spectrum. In some embodiments the nuclear fission reactor may include a traveling wave reactor, in which case the fast spectrum neutrons may be part of a nuclear fission traveling wave. In some other embodiments the nuclear fission reactor may include a fast breeder reactor, like a liquid metal fast breeder reactor or a gas-cooled fast breeder reactor, or the like. In some embodiments the neutron absorbing material 214 may be configured to reduce moderation of neutrons. For example, the neutron absorbing material 14 may have a suitably large atomic mass that can help reduce the amount of slowing down of fast spectrum neutrons. As such, a reduction may be made in softening of the neutron spectrum from the fast neutron spectrum toward neutron spectrums having neutron energy levels less than around 0.1 MeV, such as an epi-thermal neutron spectrum or a thermal neutron spectrum. It will be appreciated that, given by way of non-limiting examples, elements of the actinide series, such as without limitation uranium and thorium, present a sufficiently large atomic mass to help reduce moderation of neutrons. The fertile nuclear fission fuel material 216, that is included in the neutron absorbing material 214, can include any type of fertile nuclear fission fuel material as desired for a particular application. For example, in some embodiments the fertile nuclear fission fuel material 216 may include uranium, such as 238U. In some other embodiments the fertile nuclear fission fuel material 16 may include thorium, such as 232Th. The fertile nuclear fission fuel material 16 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. In some applications it may be desirable to soften the neutron spectrum within the fast neutron spectrum toward a softer neutron spectrum that is still within the fast neutron spectrum—that is, at least around 0.1 MeV. For example, in some applications it may be desired to soften the neutron spectrum to reduce irradiation damage. To that end and referring now to FIGS. 2B-2G, in some embodiments the reactivity control rod 212 may also include neutron moderating material 220 in addition to the fertile nuclear fission fuel material 216. The neutron moderating material 220 may include any suitable neutron moderating material as desired for a particular application. Given by way of non-limiting example, the neutron moderating material 220 may include any one or more of hydrogen, deuterium, helium, lithium, boron, carbon, graphite, sodium, lead, and the like. When the neutron moderating material 220 is provided, the neutron moderating material 220 may be distributed within the reactivity control rod 212 in any manner as desired for a particular application. For example and as shown in FIGS. 2B-2F by way of illustration and not of limitation, in some embodiments the neutron moderating material 220 may be substantially heterogeneously distributed within the reactivity control rod 212. Given by way of non-limiting examples, the neutron moderating material 220 may be heterogeneously distributed in disks 221 (FIGS. 2B and 2C). The disks 221 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 212 (as shown in FIG. 2B) or substantially transverse to the axial orientation of the reactivity control rod 212 (as shown in FIG. 2C). Given by way of further non-limiting examples, the neutron moderating material 220 may be heterogeneously distributed toward ends of the reactivity control rod 212 (as shown in FIG. 2D) or toward a middle of the reactivity control rod 212 (as shown in FIG. 2E). Given by way of a further non-limiting example, the neutron moderating material 220 may be provided as a rod follower 223 (as shown in FIG. 2F). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 2G, the neutron moderating material 220 may be substantially homogeneously distributed within the reactivity control rod 212. Referring now to FIGS. 2H-2M, in some embodiments the neutron absorbing material 214 may also include neutron absorbing poison 222 in addition to the fertile nuclear fission fuel material 216. The neutron absorbing poison 222 may include any suitable neutron absorbing poison as desired. For example and given by way of non-limiting examples, the neutron absorbing poison 222 may include any one or more of silver, indium, cadmium, gadolinium, hafnium, lithium, 3He, fission products, protactinium, neptunium, boron, and the like. The neutron absorbing poison 222 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. When the neutron absorbing poison 222 is provided, the neutron absorbing poison 222 may be distributed within the reactivity control rod 212 in any manner as desired for a particular application. For example and as shown in FIGS. 2H-2L by way of illustration and not of limitation, in some embodiments the neutron absorbing poison 222 may be substantially heterogeneously distributed within the reactivity control rod 212. Given by way of non-limiting examples, the neutron absorbing poison 222 may be heterogeneously distributed in disks 225 (FIGS. 2H and 2I). The disks 225 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 212 (as shown in FIG. 2H) or substantially transverse to the axial orientation of the reactivity control rod 212 (as shown in FIG. 2I). Given by way of further non-limiting examples, the neutron absorbing poison 222 may be heterogeneously distributed toward ends of the reactivity control rod 212 (as shown in FIG. 2J) or toward a middle of the reactivity control rod 212 (as shown in FIG. 2K). Given by way of a further non-limiting example, the neutron absorbing poison 222 may be provided as a rod follower 227 (as shown in FIG. 2L). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 2M, the neutron absorbing poison 222 may be substantially homogeneously distributed within the reactivity control rod 212. In some embodiments and referring now to FIGS. 2H-2P, an effect on reactivity achievable by the fertile nuclear fission fuel material 216 may equalized toward an effect on reactivity achievable by portions of the neutron absorbing poison 222. For example, such equalization may be desirable to mitigate localized flux peaking. It will be appreciated that such equalization may be effected regardless of whether the fertile nuclear fission fuel material 216 is distributed heterogeneously or homogeneously and regardless of whether the neutron absorbing poison 222 is distributed heterogeneously (FIGS. 2H-2L and FIGS. 2O-2P) or homogeneously (FIG. 2M). In some other embodiments and still referring to FIGS. 2H-2P, reactivity effect of the fertile nuclear fission fuel material 216 and reactivity effect of the neutron absorbing poison 222 may be locally tailored as desired for a particular application. For example, in some embodiments and as shown generally in FIG. 2N the reactivity control rod 212 has a region 224 and a region 226. It will be appreciated that the regions 224 and 226 may be located anywhere within the reactivity control rod 212 as desired. No limitation is implied, and is not to be inferred, by virtue of appearance in the drawings which are provided for illustration purposes only. A concentration 228 of the neutron absorbing poison 222 is disposed in the region 224 and a concentration 230 of the neutron absorbing poison 222 is disposed in the region 226. A concentration 232 of the fertile nuclear fission fuel material 216 is disposed in the region 224 and a concentration 234 of the fertile nuclear fission fuel material 216 is disposed in the region 226. It will be appreciated that concentration may be determined per volume basis, per area basis, or per length basis, as desired. It will be appreciated that reactivity effects of the concentrations 228 and 230 of the neutron absorbing poison 222 and reactivity effects of the concentrations 232 and 234 of the fertile nuclear fission fuel material 216 may be tailored as desired for a particular application. For example, in some embodiments and as shown in FIGS. 2H-2P a reactivity effect of the concentration 230 of the neutron absorbing poison 222 may be substantially equalized with a reactivity effect of the concentration 232 of the fertile nuclear fission fuel material 216. In some other embodiments and as also shown in FIGS. 2H-2P a reactivity effect of the concentration 228 of the neutron absorbing poison 222 may be substantially equalized with a reactivity effect of the concentration 234 of the fertile nuclear fission fuel material 216. In some other embodiments and as shown in FIGS. 2H-2P a reactivity effect of the concentration 230 of the neutron absorbing poison 222 may be different from a reactivity effect of the concentration 232 of the fertile nuclear fission fuel material 216. In other embodiments a reactivity effect of the concentration 228 of the neutron absorbing poison 222 may be different from a reactivity effect of the concentration 234 of the fertile nuclear fission fuel material 216. Other reactivity effects may be affected as desired. For example and as shown in FIGS. 2H-2P, in some embodiments a sum of reactivity effects of the concentration 228 of the neutron absorbing poison 222 and the concentration 232 of the fertile nuclear fission fuel material 216 may be substantially equalized toward a sum of reactivity effects of the concentration 230 of the neutron absorbing poison 222 and the concentration 234 of the fertile nuclear fission fuel material 216. In some other embodiments, reactivity effect is substantially constant between the region 224 and the region 226. If desired, concentration of the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may vary. For example and as shown in FIGS. 2O and 2P, in some embodiments concentration the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may change along a continuous gradient. Given by way of non-limiting example, as shown in FIG. 2O the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be provided in wedges 236 and 238, respectively, that abut each other along their hypotenuse 240. Given by way of another non-limiting example, as shown in FIG. 2P the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be provided in mated frustoconical sections 242 and 244, respectively. It will be appreciated that the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be provided in other suitable arrangements in which their concentrations change along a continuous gradient, and arrangements are not to be limited to those shown in FIGS. 2G and 2H by way of illustration and not of limitation. In some other embodiments, concentration of the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may change along a non-continuous gradient. For example, concentration of the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may change along a non-continuous gradient as a result of heterogeneous distribution as shown in FIGS. 2H-2L. In such cases, concentration of the neutron absorbing poison 222 can vary along a non-contiguous gradient because the neutron absorbing poison 222 is provided in discrete locations (as opposed to homogeneous distribution). Also in such cases, concentration of the fertile nuclear fission fuel material 216 can vary along a non-contiguous gradient because the fertile nuclear fission fuel material 216 is provided in discrete locations that are separated from each other by the discrete locations of the neutron absorbing poison 222. In some embodiments the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be spatially fixed relative to each other. That is, in such arrangements the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 do not physically move in relation to each other. However, in some other embodiments the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be spatially movable relative to each other. Given by way of non-limiting example and referring briefly to FIGS. 2H-2L and 2O-2P, any one or more of the discrete locations of the neutron absorbing poison 222, such as without limitation the disks 225, may be slidably received in the reactivity control rod 212 and may be moved in and out of the reactivity control rod 212 as desired by a suitable mechanism, such as a control rod drive mechanism or the like. Referring now to FIG. 2Q, in some embodiments the reactivity control rod 212 may define at least one chamber 256 configured to accumulate fission products. For example, when provided the chamber 256 may include a plenum 258. In some embodiments the plenum 258 may be located at least one mean free path λT for fission-inducing neutrons from the fertile nuclear fission fuel material 216. In some embodiments a backflow prevention device 260, such as a check valve like a ball check vale or the like, may be provided to help prevent re-entry into the reactivity control rod 212 of fission product gases that have outgassed from the reactivity control rod 212. As mentioned above, the actuator 217 is responsive to at least one reactivity parameter. In some embodiments, the reactivity control system 210 may include an apparatus configured to determine the reactivity parameter. Given by way of non-limiting examples and referring now to FIGS. 2R-2AL, the apparatus may include at least one sensor 218. As shown in FIGS. 2R-2AL, in some embodiments the sensor 218 may be physically associated with the reactivity control rod 210. Given by way of non-limiting examples, in FIGS. 2R-2AI, the sensor 218 may be physically associated with embodiments of the reactivity control rod 210 that have been shown and explained with reference to FIGS. 2A-2Q. In such cases, details have already been set forth regarding embodiments of the reactivity control rod 210 with reference to FIGS. 2A-2Q and need not be repeated for an understanding. In such embodiments the sensor 218 may be physically associated with the reactivity control rod 212 in any suitable physical association as desired. For example and referring to FIG. 2AI, in some embodiments physical association may include the sensor 218 being located within an interior 246 of the reactivity control rod 212. For example, the sensor 218 may be located via any suitable attachment method on an interior surface 248 of a cladding wall 250 of the reactivity control rod 212. As a further example and referring now to FIG. 2AJ, in some other embodiments physical association may include the sensor 218 being located proximate an exterior 252 of the reactivity control rod 212. For example, the sensor 218 may be located via any suitable method on an exterior surface 254 of the cladding wall 250. It will be appreciated that the sensor 218 need not be physically associated with the reactivity control rod 212. To that end, in some embodiments, the sensor 218 is not physically associated with the reactivity control rod 212. For example, in some embodiments the sensor 218 may be located at a position that is separate from the reactivity control rod 212 but that permits the sensor 218 to sense the reactivity parameter desired to be sensed. Given by way of non-limiting example, the sensor 218 may be located at a position that is separate but no more than one mean free path λT for fission-inducing neutrons from the reactivity control rod 212. Any one or more of various reactivity parameters associated with the reactivity control rod 212 may be sensed with the sensor 218. Given by way of non-limiting examples, the sensed reactivity parameter may include any one or more of parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and/or neutron spectrum. The sensor 218 may include any suitable sensor that is configured to sense the reactivity parameter that is desired to be sensed. Given by way of non-limiting example, in some embodiments the sensor 218 may include at least one fission detector, such as without limitation a micro-pocket fission detector. In some other embodiments the sensor 218 may include a neutron flux monitor, such as without limitation a fission chamber and/or an ion chamber. In some embodiments the sensor 218 may include a neutron fluence sensor, such as without limitation an integrating diamond sensor. In some embodiments the sensor 218 may include a fission product detector, such as without limitation a gas detector, a β detector, and/or a γ detector. In some embodiments, when provided, the fission product detector may be configured to measure a ratio of isotope types in fission product gas. In some embodiments the sensor 18 may include a temperature sensor. In some other embodiments the sensor 218 may include a pressure sensor. In some embodiments the sensor 218 may include a power sensor, such as without limitation a power range nuclear instrument. In some embodiments, if desired the sensor 218 may be replaceable. In some other embodiments, the reactivity parameter may be determined without being sensed by a sensor. Given by way of non-limiting example, in some embodiments the apparatus may include electrical circuitry (not shown) configured to determine at least one reactivity parameter (which have been discussed above). The reactivity parameter may be determined in any suitable manner. Given by way of non-limiting example, the reactivity parameter may be retrieved from a look-up table using operating parameters, such as temperature, pressure, power level, time in core life (as measured in effective full power hours), and the like, as entering arguments. Given by way of another non-limiting example, the reactivity parameter may be modeled, such as by running suitable neutronics modeling software on a suitable computer. Given by way of illustration, suitable neutronics modeling software includes MCNP, CINDER, REBUS, and the like. In a further non-limiting example, the reactivity parameter may be determined by a reactor operator or any other person skilled in the art based on prior knowledge or experience. In a general sense, those skilled in the art will recognize that various aspects described herein (including the electrical circuitry configured to determine at least one reactivity parameter) can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof that can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Referring to FIG. 2AK, in some embodiments a calibration device 262 configured to calibrate the sensor 218 may be provided. It will be appreciated that, when provided, the calibration device 262 suitably is a source having known characteristics or attributes of the reactivity parameter, discussed above, that is sensed by the sensor 218. Referring to FIG. 2AL, in some embodiments at least one communications device 264 may be operatively coupled to the sensor 218 as generally indicated at 266. The communications device 218 suitably is any acceptable device that can operatively couple the sensor 218 in signal communication with a suitable communications receiving device (not shown) as generally indicated at 268. Given by way of non-limiting examples, the communications device 264 may include an electrical cable, a fiber optic cable, a telemetry transmitter, a radiofrequency transmitter, an optical transmitter, or the like. Referring now to FIGS. 2A-2AL, the reactivity control rod 212 is operationally coupled, as indicated generally at 219, to the actuator 217 in any suitable manner as desired. For example, in some embodiments the reactivity control rod 212 may be electromagnetically coupled to the actuator 217. In some other embodiments the reactivity control rod 212 may be mechanically linked to the actuator 217. Referring to FIG. 2AM, in some embodiments the reactivity control system 210 may include an actuator controller 270 that is configured to generate a rod control signal 272. In such embodiments, the actuator 217 is configured to move the reactivity control rod 217 that is operationally coupled thereto (as generally indicated at 219) responsive to the rod control signal 272. The actuator controller 270 generates the rod control signal 272 and communicates the rod control signal 272 in signal communication to the actuator 217. Referring to FIG. 2AN, in some embodiments a communications device 274 is configured to communicate the rod control signal 272 from the actuator controller 270 to the actuator 217. The communications device 274 suitably is any acceptable device that can operatively couple the actuator controller 270 in signal communication with the actuator 217. Given by way of non-limiting examples, the communications device 274 may include an electrical cable, a fiber optic cable, a telemetry transmitter, a radiofrequency transmitter, an optical transmitter, or the like. The actuator controller 270 may generate the rod control signal 272 in any suitable manner as desired. For example and referring to FIG. 2AO, in some embodiments the actuator controller 270 may include an operator interface 276. Given by way of non-limiting example, in some embodiments the operator interface 276 may include a shim switch. Referring to FIG. 2AP, in some other embodiments the actuator controller 270 may include electrical circuitry 278 that is configured to automatically generate the rod control signal 272 based upon at least one reactivity parameter (which have been discussed above). Referring now to FIGS. 2A-2AP, the actuator 217 may be any suitable actuator as desired for a particular application. Given by way of non-limiting example, in some embodiments the actuator 217 may include a reactivity control rod drive mechanism. In some embodiments the actuator 217 may be configured to drive the reactivity control rod 212 bidirectionally. That is, when the reactivity control rod 212 is provided for use in a nuclear fission reactor, the reactivity control rod 212 may be driven into and/or out of a core of the nuclear fission reactor as desired. In some other embodiments, the actuator 217 may be further configured to stop driving the reactivity control rod 217 at least one intermediate position between a first stop position and a second stop position. Illustrative Nuclear Fission Traveling Wave Reactor Referring now to FIG. 3, in some embodiments an illustrative nuclear fission traveling wave reactor 300 having a fast neutron spectrum may include any illustrative embodiment of the reactivity control system 210 (FIGS. 2A-2AP). Given by way of non-limiting example, the nuclear fission traveling wave reactor 300 includes an illustrative nuclear fission reactor core 331. The nuclear fission reactor core 331 includes suitable nuclear fission fuel material 333 that is configured to propagate therein a nuclear fission traveling wave having a fast neutron spectrum. As described above, the reactivity control system 210 includes reactivity control rods 212. Each reactivity control rod 212 includes neutron absorbing material configured to absorb fast spectrum neutrons of the nuclear fission traveling wave. At least a portion of the neutron absorbing material includes fertile nuclear fission fuel material. The reactivity control system 210 also includes actuators 217. Each of the actuators 217 is responsive to at least one reactivity parameter and is operationally coupled to at least one of the reactivity control rods 212, as indicated generally at 219. In some embodiments, the reactivity parameter may include at least one reactivity parameter of the nuclear fission traveling wave reactor. However, in some other embodiments and as discussed above, the reactivity parameter may include at least one reactivity parameter of at least one of the reactivity control rods 212. In various embodiments the reactivity parameter may include one or more reactivity parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and/or neutron spectrum. It will be appreciated that the reactivity control system 210 included in the nuclear fission traveling wave reactor 300 may be embodied in any manner desired as discussed above. For example, the reactivity control system and any of its components may be embodied, without limitation, as discussed above with reference to any one or more of FIGS. 2A-2AP. Because embodiments of the reactivity control system 210 have been discussed in detail above, for sake of brevity details need not be repeated for an understanding. Illustrative details of embodiments of the nuclear fission traveling wave reactor 300 will be set forth below. It will be appreciated that the nuclear fission traveling wave reactor 300 is a non-limiting example that is set forth below for purposes of illustration and not of limitation. The nuclear fission reactor core 333 is housed within an illustrative reactor core enclosure 335 which acts to maintain vertical coolant flow through the core. In some embodiments the reactor core enclosure 335 may also function as a radiation shield to protect in-pool components such as heat exchangers and the like from neutron bombardment. The reactivity control rods 212 longitudinally extend into the nuclear fission reactor core 331 for controlling the fission process occurring therein, as discussed above. The nuclear fission reactor core 331 is disposed within an illustrative reactor vessel 337. In some embodiments the reactor vessel 337 is filled to a suitable amount (such as about 90% or so) with a pool of coolant 339, such as liquid metal like sodium, potassium, lithium, lead, mixtures thereof, and the like, or liquid metal alloys such as lead-bismuth, to such an extent that the nuclear fission reactor core 331 is submerged in the pool of coolant. Suitably, in an illustrative embodiment contemplated herein, the coolant is a liquid sodium (Na) metal or sodium metal mixture, such as sodium-potassium (Na—K). In addition, in some embodiments a containment vessel 341 sealingly surrounds parts of the nuclear fission traveling wave reactor 300. In some embodiments a primary coolant pipe 343 is coupled to the nuclear fission reactor core 331 for allowing a suitable coolant to flow through the nuclear fission reactor core 331 along a coolant flow stream or flow path 345 in order to cool the nuclear fission reactor core 331. In various embodiments the primary coolant pipe 343 may be made from materials such as, without limitation, stainless steel or from non-ferrous alloys, zirconium-based alloys, or other suitable structural materials or composites. In some embodiments the heat-bearing coolant generated by the nuclear fission reactor core 331 flows along the coolant flow path 345 to an intermediate heat exchanger 347 that is also submerged in the pool of coolant 339. The intermediate heat exchanger 347 may be made from any suitable material, such as without limitation stainless steel, that is sufficiently resistant to heat and corrosive effects of the coolant, such as without limitation liquid sodium, in the pool of coolant 339. The coolant flowing along the coolant flow path 345 flows through the intermediate heat exchanger 347 and continues through the primary coolant pipe 343. It will be appreciated that the coolant leaving intermediate heat exchanger 347 has been cooled due to heat transfer occurring in the intermediate heat exchanger 347. In some embodiments a pump 349, which may be an electro-mechanical pump or an electromagnetic pump as desired, is coupled to the primary coolant pipe 343. In such embodiments the pump 349 is in fluid communication with the coolant carried by the primary coolant pipe 343. The pump 349 pumps the coolant through the primary coolant pipe 343, through the nuclear fission reactor core 331, along the coolant flow path 345, and into the intermediate heat exchanger 347. A secondary coolant pipe 351 is provided for removing heat from the intermediate heat exchanger 347. The secondary coolant pipe 351 includes a secondary hot leg pipe segment 353 and a secondary cold leg pipe segment 355. The secondary hot leg pipe segment 353 and the secondary cold leg pipe segment 355 are integrally connected to the intermediate heat exchanger 347. The secondary coolant pipe 351 contains a secondary coolant, that is a fluid such as any one of the coolant choices previously mentioned. The secondary hot leg pipe segment 353 extends from the intermediate heat exchanger 347 to a steam generator 357. In some embodiments, if desired, the steam generator 357 may include a superheater. After passing through the steam generator 357, the secondary coolant flowing through the secondary loop pipe 351 and exiting the steam generator 357 is at a lower temperature and enthalpy than before entering the steam generator 357 due to heat transfer occurring within the steam generator 357. After passing through the steam generator 357, the secondary coolant is pumped, such as by means of a pump 359, which may be an electro-mechanical pump or an electromagnetic pump or the like, along the secondary cold leg pipe segment 355, which extends into the intermediate heat exchanger 347 for providing the previously mentioned heat transfer. Disposed in the steam generator 357 is a body of water 361 having a predetermined temperature. The secondary coolant flowing through the secondary hot leg pipe segment 353 will transfer its heat by means of conduction and convection to the body of water 361, which is at a lower temperature than the secondary coolant flowing through the secondary hot leg pipe segment 353. As the secondary coolant flowing through the secondary hot leg pipe segment 353 transfers its heat to the body of water 361, a portion of the body of water 361 will vaporize to steam 363 according to the predetermined temperature within the steam generator 357. The steam 363 will then travel through a steam line 365. One end of the steam line 365 is in vapor communication with the steam 363 and another end of the steam line 365 is in liquid communication with the body of water 361. A rotatable turbine 367 is coupled to the steam line 365 such that the turbine 367 rotates as the steam 363 passes therethrough. An electrical generator 369 is coupled to the turbine 367 by a rotatable turbine shaft 371. The electrical generator 369 generates electricity as the turbine 367 rotates. A condenser 373 is coupled to the steam line 365 and receives the steam 363 passing through the turbine 367. The condenser 373 condenses the steam 363 to liquid water and passes any waste heat via a recirculation fluid path 375 and a condensate pump 377, such as an electro-mechanical pump, to a heat sink 379, such as a cooling tower, which is associated with the condenser 373. The feed water condensed by the condenser 373 is pumped along a feed water line 381 from the condenser 373 to the steam generator 357 by a feed water pump 383, which may be an electro-mechanical pump that is interposed between the condenser 373 and the steam generator 357. Embodiments of the nuclear fission reactor core 331 may include any suitable configuration as desired to accommodate the reactivity control system 210. In this regard, in some embodiments the nuclear fission reactor core 331 may be generally cylindrically shaped to obtain a generally circular transverse cross section. In some other embodiments the nuclear fission reactor core 331 may be generally hexagonally shaped to obtain a generally hexagonal transverse cross section. In other embodiments the nuclear fission reactor core 331 may be generally parallelepiped shaped to obtain a generally rectangular transverse cross section. Regardless of the configuration or shape selected for the nuclear fission reactor core 331, the nuclear fission reactor core 331 is operated as a traveling wave nuclear fission reactor core. For example, a nuclear fission igniter (not shown for clarity), which includes an isotopic enrichment of nuclear fissionable material, such as, without limitation, U-233, U-235 or Pu-239, is suitably located in the nuclear fission reactor core 331. Neutrons are released by the igniter. The neutrons that are released by the igniter are captured by fissile and/or fertile material within the nuclear fission fuel material 333 to initiate a nuclear fission chain reaction. The igniter may be removed once the fission chain reaction becomes self-sustaining, if desired. The igniter initiates a three-dimensional, traveling wave or “burn wave”. When the igniter generates neutrons to cause “ignition”, the burn wave travels outwardly from the igniter so as to form the traveling or propagating burn wave. Speed of the traveling burn wave may be constant or non-constant. Thus, the speed at which the burn wave propagates can be controlled. For example, longitudinal movement of the reactivity control rods 210 in a predetermined or programmed manner can drive down or lower neutronic reactivity of vented nuclear fission fuel modules 30. In this manner, neutronic reactivity of nuclear fuel that is presently being burned behind the burn wave or at the location of the burn wave is driven down or lowered relative to neutronic reactivity of “unburned” nuclear fuel ahead of the burn wave. Controlling reactivity in this manner maximizes the propagation rate of the burn wave subject to operating constraints for the nuclear fission reactor core 331, such as amount of permissible fission product production and/or neutron fluence limitations of reactor core structural materials. The basic principles of such a traveling wave nuclear fission reactor are disclosed in more detail in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; and U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the entire contents of which are hereby incorporated by reference. It will be appreciated that the embodiment of the nuclear fission traveling wave reactor 300 described above is set forth as a non-limiting example for purposes of illustration only and not of limitation. In some other embodiments, the nuclear fission traveling wave reactor 300 may be a gas-cooled fast nuclear fission traveling wave reactor that includes a suitable gas coolant, such as helium or the like. In such an embodiment, a gas-driven turbine-generator may be driven by the gas coolant. Illustrative Methods, Systems, and Computer Software Program Products Following are a series of flowcharts depicting implementations of processes. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an overall “big picture” viewpoint and thereafter the following flowcharts present alternate implementations and/or expansions of the “big picture” flowcharts as either sub-steps or additional steps building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an overall view and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular and/or object-oriented program design paradigms. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Referring now to FIG. 4A, a method 400 is provided for controlling reactivity in a nuclear fission reactor having a fast neutron spectrum. The method 400 starts at a block 402. At a block 404 a desired reactivity parameter within a selected portion of the nuclear fission reactor having a fast neutron spectrum is determined. At a block 406 at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, is adjusted responsive to the desired reactivity parameter. The method 400 stops at a block 408. It will be appreciated that the method 400 may be performed with respect to any nuclear fission reactor having a fast neutron spectrum. In some embodiments, the method 400 may be performed with respect to a nuclear fission traveling wave reactor, in which case the fast spectrum neutrons may be part of a nuclear fission traveling wave. In some other embodiments, the method 400 may be performed with respect to any suitable fast breeder reactor, such as a liquid metal fast breeder reactor, a gas-cooled fast breeder reactor, or the like. Thus, no limitation to any particular type of nuclear fission reactor having a fast neutron spectrum is intended and should not be inferred. Illustrative details will be set forth below by way of non-limiting examples. In various embodiments the desired reactivity parameter may be determined with respect to any portion of a nuclear fission reactor as desired. For example and referring to FIG. 4B, in some embodiments determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum at the block 404 may include determining at least one desired reactivity parameter of the fertile nuclear fission fuel material at a block 410. In some other embodiments and referring to FIG. 4C, determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum at the block 404 may include determining at least one desired reactivity parameter of the at least one reactivity control rod at a block 412. In some other embodiments and referring to FIG. 4D, determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum at the block 404 may include determining at least one desired reactivity parameter of the nuclear fission reactor at a block 414. In some embodiments the reactivity control rod may be adjusted responsive to a difference between the desired reactivity parameter and a determination of the reactivity parameter. For example and referring to FIGS. 4A and 4E, in some embodiments at a block 416 at least one determined reactivity parameter may be determined. Referring additionally to FIG. 4F, in some embodiments at a block 418 a difference between the desired reactivity parameter and the at least one determined reactivity parameter may be determined. Referring additionally to FIG. 4G, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 406 may include adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the difference between the desired reactivity parameter and the at least one determined reactivity parameter at a block 420. The determined reactivity parameter may be determined in any suitable manner as desired. For example and referring now to FIGS. 4E and 4H, in some embodiments determining at least one determined reactivity parameter at the block 416 may include predicting at least one reactivity parameter at a block 422. Referring to FIGS. 4E and 4I, in some embodiments determining at least one determined reactivity parameter at the block 416 may include modeling at least one reactivity parameter at a block 424. Referring to FIGS. 4E and 4J, in some embodiments determining at least one determined reactivity parameter at the block 416 may include selecting at least one predetermined reactivity parameter at a block 426. Referring to FIGS. 4E and 4K, in some other embodiments determining at least one determined reactivity parameter at the block 416 may include sensing at least one reactivity parameter at a block 428. It will be appreciated that any desired reactivity parameter may be sensed at the block 428 in any suitable manner. For example and referring to FIGS. 4K and 4L, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing a time history of at least one reactivity parameter at a block 430. Sensing a time history may be performed as desired, such as by sensing and recording or storing the sensed reactivity parameter more than one time. Given by way of non-limiting examples, a time history of at least one reactivity parameter may include, without limitation, a rate of the reactivity parameter, accumulation of the reactivity parameter, total fissions, or the like. Referring to FIGS. 4K and 4M, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing at least one radioactive decay event at a block 432. Referring to FIGS. 4K and 4N, in some embodiments sensing at least one reactivity parameter at the block 428 may include detecting fission at a block 434. Referring to FIGS. 4K and 4O, in some embodiments sensing at least one reactivity parameter at the block 428 may include monitoring neutron flux at a block 436. Referring to FIGS. 4K and 4P, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing neutron fluence at a block 438. Referring to FIGS. 4K and 4Q, in some embodiments sensing at least one reactivity parameter at the block 428 may include detecting fission products at a block 440. Referring to FIGS. 4K and 4R, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing temperature at a block 442. Referring to FIGS. 4K and 4S, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing pressure at a block 444. Referring to FIGS. 4K and 4T, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing power level at a block 446. Referring now to FIGS. 4A and 4U, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 406 may include moving, in at least one of two directions, at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at a block 448. In various embodiments the directions may include axial directions in the nuclear fission reactor, radial directions in the nuclear fission reactor, and/or lateral directions in the nuclear fission reactor. Referring now to FIGS. 4A and 4V, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing a difference in reactivity in association with a change in position of the reactivity control rod at a block 450. Referring to FIGS. 4A and 4W, in some embodiments a sensor that is configured to sense at least one reactivity parameter may be calibrated at a block 452. Referring now to FIG. 5A, a method 500 is provided for operating a nuclear fission traveling wave reactor having a fast neutron spectrum. The method 500 starts at a block 502. At a block 503 a nuclear fission traveling wave having a fast neutron spectrum is propagated in a nuclear fission traveling wave reactor core. At a block 504 a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor is determined. At a block 506 at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, is adjusted responsive to the desired reactivity parameter. The method 500 stops at a block 508. Illustrative details will be set forth below by way of non-limiting examples. Referring now to FIGS. 5A and 5B, in some embodiments a nuclear fission traveling wave having a fast neutron spectrum may be initiated in the nuclear fission traveling wave reactor core at a block 509. In various embodiments the desired reactivity parameter may be determined with respect to any portion of the nuclear fission traveling wave reactor as desired. For example and referring to FIG. 5C, in some embodiments determining a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor at the block 504 may include determining at least one desired reactivity parameter of the fertile nuclear fission fuel material at a block 510. In some other embodiments and referring to FIG. 5D, determining a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor at the block 504 may include determining at least one desired reactivity parameter of the at least one reactivity control rod at a block 512. In some other embodiments and referring to FIG. 5E, determining a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor at the block 504 may include determining at least one desired reactivity parameter of the nuclear fission traveling wave reactor at a block 514. In some embodiments the reactivity control rod may be adjusted responsive to a difference between the desired reactivity parameter and a determination of the reactivity parameter. For example and referring to FIGS. 5A and 5F, in some embodiments at a block 516 at least one determined reactivity parameter may be determined. Referring additionally to FIG. 5G, in some embodiments at a block 518 a difference between the desired reactivity parameter and the at least one determined reactivity parameter may be determined. Referring additionally to FIG. 5H, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 506 may include adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the difference between the desired reactivity parameter and the at least one determined reactivity parameter at a block 520. The determined reactivity parameter may be determined in any suitable manner as desired. For example and referring now to FIGS. 5F and 5O, in some embodiments determining at least one determined reactivity parameter at the block 516 may include predicting at least one reactivity parameter at a block 522. Referring to FIGS. 5F and 5J, in some embodiments determining at least one determined reactivity parameter at the block 516 may include modeling at least one reactivity parameter at a block 524. Referring to FIGS. 5F and 5K, in some embodiments determining at least one determined reactivity parameter at the block 516 may include selecting at least one predetermined reactivity parameter at a block 526. Referring to FIGS. 5F and 5L, in some other embodiments determining at least one determined reactivity parameter at the block 516 may include sensing at least one reactivity parameter at a block 528. It will be appreciated that any desired reactivity parameter may be sensed at the block 528 in any suitable manner. For example and referring to FIGS. 5L and 5M, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing a time history of at least one reactivity parameter at a block 530. Sensing a time history may be performed as desired, such as by sensing and recording or storing the sensed reactivity parameter more than one time. Given by way of non-limiting examples, a time history of at least one reactivity parameter may include, without limitation, a rate of the reactivity parameter, accumulation of the reactivity parameter, total fissions, or the like. Referring to FIGS. 5L and 5N, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing at least one radioactive decay event at a block 532. Referring to FIGS. 5L and 5O, in some embodiments sensing at least one reactivity parameter at the block 528 may include detecting fission at a block 534. Referring to FIGS. 5L and 5P, in some embodiments sensing at least one reactivity parameter at the block 528 may include monitoring neutron flux at a block 536. Referring to FIGS. 5L and 5Q, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing neutron fluence at a block 538. Referring to FIGS. 5L and 5R, in some embodiments sensing at least one reactivity parameter at the block 528 may include detecting fission products at a block 540. Referring to FIGS. 5L and 5S, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing temperature at a block 542. Referring to FIGS. 5L and 5T, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing pressure at a block 544. Referring to FIGS. 5L and 5U, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing power level at a block 546. Referring now to FIGS. 5A and 5V, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 506 may include moving, in at least one of two directions, at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at a block 548. In various embodiments the directions may include axial directions in the nuclear fission traveling wave reactor, radial directions in the nuclear fission traveling wave reactor, and/or lateral directions in the nuclear fission traveling wave reactor. Referring now to FIGS. 5A and 5W, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing a difference in reactivity in association with a change in position of the reactivity control rod at a block 550. Referring to FIGS. 5A and 5X, in some embodiments a sensor that is configured to sense at least one reactivity parameter may be calibrated at a block 552. Referring now to FIG. 6A, an illustrative system 610 is provided for controlling reactivity in a nuclear fission reactor (not shown) having a fast neutron spectrum. The system 610 includes means 612 for determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum. The system 610 also includes means 614 for adjusting at least one reactivity control rod (not shown) having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter. In various embodiments the determining means 612 may include suitable electrical circuitry. As discussed above, various aspects described herein (including the means 612 for determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum) can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof that can be viewed as being composed of various types of “electrical circuitry.” Consequently, it is emphasized that, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. In various embodiments the adjusting means 614 may include any suitable electro-mechanical system, such as without limitation an actuator. Given by way of illustration and not limitation, a non-limiting example of an actuator includes a control rod drive mechanism. However, it will be appreciated that, in a general sense, the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, and electro-magnetically actuated devices, or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment), and any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, as well as other systems such as motorized transport systems, factory automation systems, security systems, and communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In some embodiments the fast spectrum neutrons may be part of a nuclear fission traveling wave. In such cases, the nuclear fission reactor may include a nuclear fission traveling wave reactor. However, it will be appreciated that in other embodiments the fast spectrum neutrons need not be part of a nuclear fission traveling wave. Thus, in some embodiments, the nuclear fission reactor may include any suitable nuclear fission reactor having a fast neutron spectrum. Referring to FIG. 6B, in some embodiments the means 612 for determining a desired reactivity parameter may include means 616 for determining at least one desired reactivity parameter of the fertile nuclear fission fuel material. In some other embodiments and referring to FIG. 6C, the means 612 for determining a desired reactivity parameter may include means 618 for determining at least one desired reactivity parameter of the at least one reactivity control rod. In some other embodiments and referring to FIG. 6D, the means 612 for determining a desired reactivity parameter may include means 620 for determining at least one desired reactivity parameter of the nuclear fission reactor. The means 616, 618, and 620 may include suitable electrical circuitry, as described above. Referring now to FIG. 6E, in some embodiments the system 610 may also include means 622 for determining at least one determined reactivity parameter. In some embodiments the means 622 may include suitable electrical circuitry, as described above. Some other embodiments of the means 622 will be discussed below. Referring now to FIG. 6F, in some embodiments the system 610 may also include means 624 for determining a difference between the desired reactivity parameter and the at least one determined reactivity parameter. The means 624 may include suitable electrical circuitry, as described above, such as without limitation a comparator. Referring additionally to FIG. 6G, in some embodiments the adjusting means 614 may include means 626 for adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, responsive to the difference between the desired reactivity parameter and the at least one determined reactivity parameter. The means 626 may include any suitable electro-mechanical system as described above, such as without limitation a control rod drive mechanism. In various embodiments the determining means 622 may determine a determined reactivity parameter in any manner as desired for a particular application. For example and referring to FIG. 6H, in some embodiments the means 622 for determining at least one determined reactivity parameter may include means 628 for predicting at least one reactivity parameter. The means 628 may include suitable electrical circuitry, as described above. Given by way of non-limiting example, the predetermined reactivity parameter may be retrieved from a look-up table using operating parameters, such as temperature, pressure, power level, time in core life (as measured in effective full power hours), and the like, as entering arguments. Referring to FIG. 6I, in some other embodiments the means 622 for determining at least one determined reactivity parameter may include means 630 for modeling at least one reactivity parameter. The means 630 may include suitable electrical circuitry, as described above, such as without limitation a suitable computer. The means 630 may also include suitable neutronics modeling software that runs on the electrical circuitry. Given by way of illustration, suitable neutronics modeling software includes MCNP, CINDER, REBUS, and the like. Referring to FIG. 6J, in some embodiments the means 622 for determining at least one determined reactivity parameter may include means 632 for selecting at least one predetermined reactivity parameter. The means 632 may include suitable electrical circuitry, as described above. Referring to FIG. 6K, in some embodiments the means 622 for determining at least one determined reactivity parameter may include means 634 for sensing at least one reactivity parameter. In various embodiments, the sensing means 634 may include any one or more of various sensors and detectors as desired for a particular purpose, as will be discussed below. Referring to FIG. 6L, in some embodiments the sensing means 634 may include means 636 for sensing a time history of at least one reactivity parameter. Sensing a time history may be performed as desired, such as by sensing and recording or storing the sensed reactivity parameter more than one time. Given by way of non-limiting examples, a time history of at least one reactivity parameter may include, without limitation, a rate of the reactivity parameter, accumulation of the reactivity parameter, total fissions, or the like. In various embodiments the means 636 may include suitable storage, such as computer memory media or computer memory storage or the like, configured to store values of the reactivity parameter over time. Referring to FIG. 6M, in some other embodiments the sensing means 634 may include 638 means for sensing at least one radioactive decay event. Given by way of non-limiting examples, the means 638 may include any one or more of suitable sensors or detectors for sensing α, β, and/or γ radiation as desired. Referring back to FIG. 6K, in various embodiments the sensing means 634 may include any suitable sensor as desired for a particular application. Given by way of illustrative examples and without limitation, in various embodiments the sensing means 634 may include any one or more sensor, such as at least one fission detector, a neutron flux monitor, a neutron fluence sensor, a fission product detector, a temperature sensor, a pressure sensor, and/or a power level sensor. Referring to FIG. 6N, in some embodiments the adjusting means 614 may include means 640 for moving, in at least one of two directions, at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter. In some embodiments, the means 640 may include an actuator such as a control rod drive mechanism and/or a rod handling system. In various embodiments, the directions may include any one or more of axial directions in the nuclear fission reactor, radial directions in the nuclear fission reactor, and/or lateral directions in the nuclear fission reactor. Referring to FIG. 6O, in some embodiments the sensing means 634 may include means 642 for sensing a difference in reactivity in association with a change in position of the reactivity control rod. In various embodiments, the means 642 may include electrical circuitry, as described above. In some embodiments the electrical circuitry may implement a suitable comparator. Referring to FIG. 6P, in some embodiments the system 610 may also include means 644 for calibrating the sensing means 634. In various embodiments the calibration means 644 suitably includes a source having known characteristics or attributes of the reactivity parameter, discussed above, that is sensed by the sensing means 634. Referring now to FIG. 7A, a method 700 is provided for determining an application of a controllably movable rod. The method 700 starts at a block 702. At a block 704 at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor is determined, the controllably movable rod including fertile nuclear fission fuel material. At a block 706 an application of the controllably movable rod, chosen from a reactivity control rod and a nuclear fission fuel rod, is determined. The method 700 stops at a block 708. In various embodiments, the application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to the at least one determined reactivity parameter in the controllably movable rod. Non limiting examples given by way of illustration and not of limitation will be described below. Referring to FIG. 7B, in some embodiments at a block 710 the determined reactivity parameter and a target reactivity parameter may be compared. In some embodiments, an application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to comparison of the determined reactivity parameter and the target reactivity parameter. Referring back to FIG. 7A, in some embodiments the at least one reactivity parameter may include a neutron absorption coefficient. Referring to FIG. 7C, in some embodiments at a block 712 the determined neutron absorption coefficient a target neutron absorption coefficient may be compared. In some embodiments, an application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to comparison of the determined neutron absorption coefficient and the target neutron absorption coefficient. For example, a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron absorption coefficient is at least the target neutron absorption coefficient. As another example, a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron absorption coefficient is less than the target neutron absorption coefficient. Referring back to FIG. 7A, in some other embodiments the at least one reactivity parameter may include a neutron production coefficient. Referring to FIG. 7D, in some embodiments at a block 714 the determined neutron production coefficient and a target neutron production coefficient may be compared. In some embodiments, an application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to comparison of the determined neutron production coefficient and the target neutron production coefficient. For example, a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron production coefficient is at least the target neutron production coefficient. As another example, a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron production coefficient is less than the target neutron production coefficient. Referring back to FIG. 7A, the at least one reactivity parameter may be determined in any manner as desired for a particular application. Given by way of non-limiting examples, determining at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor may be based on neutron exposure history of the controllably movable rod, a property of fertile nuclear fission fuel material of the controllably movable rod, a property of fissile nuclear fission fuel material of the controllably movable rod, a property of neutron absorbing poison of the controllably movable rod, and/or a property of fission products of the controllably movable rod. Referring to FIG. 7E, in some embodiments determining at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at the block 704 may include monitoring at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at a block 716. Referring to FIG. 7F, in some other embodiments determining at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at the block 704 may include predicting at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at a block 718. Referring to FIG. 7G, in some embodiments predicting at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at the block 718 may include calculating at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at a block 720. Referring now to FIG. 8A, a system 810 is provided for determining an application of a controllably movable rod. An apparatus 812 is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor, the controllably movable rod including fertile nuclear fission fuel material. Electrical circuitry 814 is configured to determine an application of the controllably movable rod chosen from a reactivity control rod and a nuclear fission fuel rod. In various embodiments, the application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to the at least one determined reactivity parameter in the controllably movable rod. Non limiting examples given by way of illustration and not of limitation will be described below. Referring to FIG. 8B, a comparator 816 may be configured to compare the determined reactivity parameter and a target reactivity parameter. In such a case, the electrical circuitry 814 may be responsive to the comparator 816. Still referring to FIG. 8B, in some embodiments the at least one reactivity parameter may include a neutron absorption coefficient. In such cases, the comparator 816 may be further configured to compare the determined neutron absorption coefficient with a target neutron absorption coefficient. The electrical circuitry 814 may be responsive to comparison of the determined neutron absorption coefficient and the target neutron absorption coefficient by the comparator 816. In some embodiments a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron absorption coefficient is at least the target neutron absorption coefficient. In some other embodiments a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron absorption coefficient is less than the target neutron absorption coefficient. Still referring to FIG. 8B, in some other embodiments the at least one reactivity parameter may include a neutron production coefficient. In such cases, the comparator 816 may be further configured to compare the determined neutron production coefficient with a target neutron production coefficient. The electrical circuitry 814 may be responsive to comparison of the determined neutron production coefficient and the target neutron production coefficient by the comparator 816. In some embodiments a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron production coefficient is at least the target neutron production coefficient. In some other embodiments a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron production coefficient is less than the target neutron production coefficient. Referring back to FIG. 8A, in various embodiments the apparatus 812 may be configured as desired to determine the reactivity parameter. For example and referring to FIG. 8C, in some embodiments the apparatus 812 may include electrical circuitry 818 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on neutron exposure history of the controllably movable rod. Referring to FIG. 8D, in some other embodiments the apparatus 812 may include electrical circuitry 820 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of fertile nuclear fission fuel material of the controllably movable rod. Referring to FIG. 8E, in some embodiments the apparatus 812 may include electrical circuitry 822 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of fissile nuclear fission fuel material of the controllably movable rod. Referring to FIG. 8F, in some embodiments the apparatus 812 may include electrical circuitry 824 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of neutron absorbing poison of the controllably movable rod. Referring to FIG. 8G, in some embodiments the apparatus 812 may include electrical circuitry 826 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of fission products of the controllably movable rod. Referring to FIG. 8H, in some embodiments the apparatus 812 may include at least one monitoring device 828 that is configured to monitor at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor. Given by way of non-limiting examples, the monitoring device 828 may include any one or more of the sensors and/or detectors described above. Referring to FIG. 8I, in some embodiments the apparatus 812 may include electrical circuitry 830 that is configured to predict at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor. For example, in some embodiments the electrical circuitry 830 may be further configured to calculate at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor. Referring to FIG. 9A, an illustrative method 900 is provided for operating a nuclear fission traveling wave reactor. The method 900 starts at a block 902. At a block 904 a reactivity control apparatus having a first worth is inserted into a first location of a reactor core of a nuclear fission traveling wave reactor. At a block 906, worth of the reactivity control apparatus is modified. At a block 908 the reactivity control apparatus is moved from the first location to a second location of the reactor core of the nuclear fission traveling wave reactor such that the reactivity control apparatus has a second worth that is different from the first worth. The method 900 stops at a block 910. Referring to FIG. 9B, in some embodiments modifying worth of the reactivity control apparatus at the block 906 may include absorbing neutrons by the reactivity control apparatus at a block 912. Referring to FIG. 9C, in some embodiments, absorbing neutrons by the reactivity control apparatus at the block 912 may include absorbing neutrons by fertile nuclear fission fuel material of the reactivity control apparatus at a block 914. In some of the cases, the second worth may be greater than the first worth. Referring to FIG. 9D, in some other embodiments modifying worth of the reactivity control apparatus at the block 906 may include modifying absorptive effect of self-shielded burnable poison of the reactivity control rod at a block 916. Referring to FIG. 9E, in some embodiments modifying absorptive effect of self-shielded burnable poison of the reactivity control rod at the block 916 may include modifying self-shielding effect of the self-shielded burnable poison at a block 918. Referring to FIG. 9F, in some embodiments modifying self-shielding effect of the self-shielded burnable poison at the block 918 may include modifying exposure of the self-shielded burnable poison to a neutron flux at a block 920. Referring to FIG. 9G, in some embodiments modifying exposure of the self-shielded burnable poison to a neutron flux at the block 920 may include modifying neutron energy at a block 922. In some embodiments the second worth may be less than the first worth. In some other embodiments the second worth may be greater than the first worth. In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, and electro-magnetically actuated devices, or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment), and any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, as well as other systems such as motorized transport systems, factory automation systems, security systems, and communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware and software implementations of aspects of systems; the use of hardware or software is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. The foregoing detailed description has set forth various embodiments of the devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In one embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link, etc.). It will be appreciated that each block of block diagrams and flowcharts, and combinations of blocks in block diagrams and flowcharts, can be implemented by computer program instructions. These computer program instructions may be loaded onto a computer or other programmable apparatus to produce a machine, such that the instructions which execute on the computer or other programmable apparatus create computer-readable media software program code configured to implement the functions specified in the block diagram or flowchart block(s). These computer program instructions may also be stored in a computer-readable memory that can direct a computer or other programmable apparatus to function in a particular manner, such that the instructions stored in the computer-readable memory produce an article of manufacture including computer-readable media software program code instructions which implement the function specified in the block diagram or flowchart block(s). The computer-readable media software program code instructions may also be loaded onto a computer or other programmable apparatus to cause a series of operational steps to be performed on the computer or other programmable apparatus to produce a computer implemented process such that the instructions which execute on the computer or other programmable apparatus provide steps for implementing the functions specified in the block diagram or flowchart block(s). Accordingly, blocks of the block diagrams or flowcharts support combinations of means for performing the specified functions, combinations of steps for performing the specified functions, and computer-readable media software program code for performing the specified functions. It will also be understood that each block of the block diagrams or flowcharts, and combinations of blocks in the block diagrams or flowcharts, can be implemented by special purpose hardware-based computer systems which perform the specified functions or steps, or combinations of special purpose hardware and computer instructions. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures can be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled”, to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable”, to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components and/or wirelessly interactable and/or wirelessly interacting components and/or logically interacting and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to.” Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, etc. unless context requires otherwise. In some instances, one or more components may be referred to herein as “configured to.” Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. Furthermore, it is to be understood that the invention is defined by the appended claims. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to inventions containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that virtually any disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms. For example, the phrase “A or B” will be understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. With respect to context, even terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Those skilled in the art will appreciate that the foregoing specific illustrative processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. One skilled in the art will recognize that the herein described components (e.g., process blocks), devices, and objects and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are within the skill of those in the art. Consequently, as used herein, the specific examples set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific example herein is also intended to be representative of its class, and the non-inclusion of such specific components (e.g., process blocks), devices, and objects herein should not be taken as indicating that limitation is desired. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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description | The present invention relates to a nuclear power plant control system and a nuclear power plant control method, and particularly relates to a nuclear power plant control system and a nuclear power plant control method capable of achieving high reliability while connecting systems having different multiplicities. A nuclear power plant control system that controls a nuclear power plant includes a plurality of independently operating systems to achieve high reliability (for example, see Patent Literature 1). An example of a configuration of the nuclear power plant control system including a plurality of systems will be described with reference to FIGS. 5 and 6. FIG. 5 is a diagram illustrating an example of a configuration of a conventional nuclear power plant control system. A nuclear power plant control system 1 illustrated in FIG. 5 is a control system that controls a nuclear reactor trip, and includes a duplex channel and a duplex train. Herein, a channel is a system that detects a phenomenon occurring in a nuclear power plant, and a train is a system that performs a logical operation based on a detection result of the channel. In the nuclear power plant control system 1, a detection unit 10a included in a channel I includes a sensor that detects a specific phenomenon (for example, a rise in pressure in a particular part and the like) occurring in the nuclear power plant, and a threshold operation unit that performs a threshold operation of a detected value of the sensor. When a detected value of the sensor exceeds a threshold, the detection unit 10a outputs a signal that indicates an occurrence of a specific phenomenon to an input unit 21a included in a train A. A detection unit 10b included in a channel II includes a sensor and a threshold operation unit similar to those of the detection unit 10a, and outputs a signal that indicates an occurrence of a specific phenomenon to an input unit 21b included in a train B when a detected value of the sensor exceeds a threshold. A trip control device 20 including the input unit 21a and the input unit 21b trips a nuclear reactor when a signal is input to at least one of the input unit 21a and the input unit 21b. By such a configuration, the nuclear power plant control system 1 may maintain a function even when a malfunction occurs in a portion of the channel or the train. FIG. 6 is a diagram illustrating an example of a configuration of a recent nuclear power plant control system. A nuclear power plant control system 2 illustrated in FIG. 6 is a control system that controls a nuclear reactor trip, and includes a quadruple channel of a channel I to a channel IV. A channel I includes a detection unit 30a and an input unit 41a. A channel II includes a detection unit 30b and an input unit 41b. A channel III includes a detection unit 30c and an input unit 41c. A channel IV includes a detection unit 30d and an input unit 41d. In the nuclear power plant control system 2, the respective detection units 30a to 30d include a sensor that detects a specific phenomenon occurring in a nuclear power plant, and a threshold operation unit that performs a threshold operation of a detected value of the sensor. When a detected value of the sensor exceeds a threshold, the detection units 30a to 30d output a signal that indicates an occurrence of a specific phenomenon to an input unit 41 included in the same channel. A trip control device 40 including input units 41a to 41d trips a nuclear reactor when a signal is input to at least two of the input units 41a to 41d. By such a configuration, the nuclear power plant control system 2 may maintain a function even when a malfunction occurs in a portion of the channel. Patent Literature 1: Japanese Patent Application Laid-open No. 6-27293 As previously described, a nuclear power plant control system includes a plurality of systems, and a multiplicity varies according to a system. In general, a recent system has a high multiplicity. For this reason, when a portion of an existing nuclear power plant control system is to be replaced by a recent equipment to improve a function and the like, a simple replacement of the equipment may not respond since a multiplicity is different. For example, the detection units 10a and 10b illustrated in FIG. 5 are to be replaced by the detection units 30a to 30d illustrated in FIG. 6, these may not be simply connected since the input units 21a and 21b are a duplex unit while the detection units 30a to 30d are a quadruple unit. In addition, when systems having different multiplicities are inappropriately connected to each other, redundancy is lost, and reliability which is significant for a nuclear power plant control system may be degraded. The invention is conceived in view of the above, and an object of the invention is to provide a nuclear power plant control system and a nuclear power plant control method capable of achieving high reliability while connecting systems having different multiplicities to each other. According to an aspect of the present invention, a nuclear power plant control system includes: detection units that detect a phenomenon occurring in a nuclear power plant for each of M systems; a start unit that starts processing corresponding to the phenomenon in a case where a signal that indicates an occurrence of the phenomenon is input from a predetermined number or more of signal lines out of L signal lines; and a majority circuit, provided for each of the L signal lines, each of which outputting a signal that indicates an occurrence of the phenomenon to a corresponding signal line in a case where the phenomenon is detected in N or more systems out of the M systems of the detection units. L is an integer greater than or equal to 1, M is an integer greater than or equal to 2, and N is an integer greater than or equal to 1. The nuclear power plant control system includes majority circuits corresponding to systems of a connection destination on a one-to-one basis, and each majority circuit outputs a signal to the connection destination in response to a detection status at a connection source. Accordingly, it is possible to achieve high reliability while connecting systems having different multiplicities to each other. Advantageously, in the nuclear power plant control system, the majority circuit is combined with a relay or a breaker provided for each of the M systems. In this aspect, since the majority circuit is combined with the relay or the breaker which is a device that reliability is verified, reliability of the majority circuit may be ensured. Advantageously, in the nuclear power plant control system, the relay or the breaker is multiplexed for each of the M systems. In this aspect, reliability of the majority circuit may be enhanced by multiplexing the relay or the breaker. According to another aspect of the present invention, a nuclear power plant control method of controlling a transmission of a signal between detection units that detect a phenomenon occurring in a nuclear power plant for each of M systems and a start unit that starts processing corresponding to the phenomenon in a case where a signal that indicates an occurrence of the phenomenon is input from a predetermined number or more of signal lines out of L signal lines, includes: receiving, by a majority circuit provided for each of the L signal lines, a signal that indicates whether the phenomenon is detected from each of the M systems, and outputting, by the majority circuit, a signal that indicates an occurrence of the phenomenon to a corresponding signal line in a case where a signal indicating that the phenomenon is detected is received from N or more systems out of the M systems. L is an integer greater than or equal to 1, M is an integer greater than or equal to 2, and N is an integer greater than or equal to 1. In the nuclear power plant control method, majority circuits corresponding to systems of a connection destination on a one-to-one basis are provided, and each majority circuit outputs a signal to the connection destination in response to a detection status at a connection source. Accordingly, it is possible to achieve high reliability while connecting systems having different multiplicities to each other. The nuclear power plant control system and the nuclear power plant control method according to the invention have an effect of achieving high reliability while connecting systems having different multiplicities to each other. Hereinafter, Embodiment of a nuclear power plant control system and a nuclear power plant control method according to the invention will be described in detail based on drawings. It should be noted that the invention is not limited to the Embodiment. In addition, components in the Embodiment contain the equivalent including a component easily assumed by those skilled in the art, and the substantially same component. First, a configuration of a nuclear power plant control system according to the Embodiment will be described with reference to FIG. 1. FIG. 1 is a diagram illustrating a schematic configuration of a nuclear power plant control system according to the Embodiment. A nuclear power plant control system 3 illustrated in FIG. 1 is a control system that controls a nuclear reactor trip, and includes detection units 30a to 30d, a trip control device 20, and majority circuits 50a and 50b. Similarly to the detection units 30a to 30d illustrated in FIG. 6, the respective detection units 30a to 30d include a sensor that detects a specific phenomenon occurring in a nuclear power plant, and a threshold operation unit that performs a threshold operation of a detected value of the sensor, and output a signal that indicates an occurrence of a specific phenomenon when a detected value of the sensor exceeds a threshold. Similarly to the trip control device 20 illustrated in FIG. 5, the trip control device 20 includes an input unit 21a and an input unit 21b, and trips a nuclear reactor when a signal that indicates an occurrence of a specific phenomenon is input to at least one of the input unit 21a and the input unit 21b. In this way, the nuclear power plant control system 3 includes a quadruple channel and a duplex train. The majority circuits 50a and 50b connect the quadruple channel to the duplex train. Specifically, the majority circuit 50a is connected to the respective detection units 30a to 30d, and outputs a signal that indicates an occurrence of a specific phenomenon to the input unit 21a through a signal line Sa when the signal that indicates an occurrence of a specific phenomenon is input from at least two of the detection units 30a to 30d. In addition, the majority circuit 50b is connected to the respective detection units 30a to 30d, and outputs a signal that indicates an occurrence of a specific phenomenon to the input unit 21b through a signal line Sb when the signal that indicates an occurrence of a specific phenomenon is input from at least two of the detection units 30a to 30d. To connect the quadruple channel to the duplex train, for example, a first channel and a second channel may be connected to a first train via an OR circuit, and a third channel and a fourth channel may be connected to a second train via an OR circuit. However, in this case, when one of the OR circuits breaks down, or a function is suspended to conduct a test, a detection result in a channel connected to the OR circuit is not reflected on a control, and reliability is significantly degraded. In addition, to connect the quadruple channel to the duplex train, for example, all channels may be connected to the first train via an OR circuit, and all channels may be connected to the second train via an OR circuit. However, in this case, when one channel merely erroneously outputs a signal, processing of a nuclear reactor trip and the like is erroneously executed even when the other channels normally operate. As illustrated in FIG. 1, when majority circuits and trains are connected to each other on a one-to-one basis by providing the same number of the majority circuits as that of the trains, and outputs of all channels are input to the respective majority circuits, it is possible to connect channels and trains having different multiplicities to each other while retaining high reliability. Next, an operation and a configuration of the majority circuits 50a and 50b illustrated in FIG. 1 will be described in detail with reference to FIGS. 2 to 4. The majority circuits 50a and 50b have similar configurations. Thus, hereinafter, the operation and the configuration will be described using the majority circuit 50a as an example. FIG. 2 is a flowchart illustrating an operation of the majority circuit 50a. The majority circuit 50a executes the operation illustrated in FIG. 2 at predetermined intervals. As illustrated in FIG. 2, in step S10, the majority circuit 50a verifies an output of a detection unit 30 of each channel. Then, when an occurrence of a specific phenomenon is detected in the detection units 30 of at least two channels (Yes in step S11), the majority circuit 50a outputs a signal that indicates the occurrence of a specific phenomenon (for example, “1”) to the input unit 21a via the signal line Sa in step S12. On the other hand, when an occurrence of a specific phenomenon is not detected in the detection units 30 of at least two channels (No in step S11), the majority circuit 50a outputs a signal indicating that the specific phenomenon does not occur (for example, “0”) to the input unit 21a via the signal line Sa in step S13. FIG. 3 is a diagram illustrating an example of a configuration of the majority circuit 50a. In the example illustrated in FIG. 3, the majority circuit 50a includes relays 51 and 52 operating based on a signal received from the channel I, relays 53 and 55 operating based on a signal received from the channel II, relays 54 and 56 operating based on a signal received from the channel III, and relays 57 and 58 operating based on a signal received from the channel IV. In this way, when the majority circuit 50a is combined with a relay which has been used for a long time, and is a device that reliability is verified, reliability of the majority circuit 50a may be ensured. In addition, by using the relay, as illustrated in FIG. 4, the channel and the train may be physically separated from each other in a portion of a contact point. When the channel and the train are desired to be separated from each other to ensure safety, it is important to physically separate the channel and the train from each other. The majority circuits 50a and 50b may be combined with a breaker which is a device that reliability is verified similarly to the relay. As described in the foregoing, in the Embodiment, since the majority circuits 50a and 50b are provided, it is possible to achieve a nuclear power plant control system having high reliability while connecting the channel and the train having different multiplicities to each other. The configuration of the nuclear power plant control system described in the above Embodiment may be arbitrarily changed without departing from the scope of the invention. For example, in the above Embodiment, the nuclear power plant control system that performs a control of a nuclear reactor trip is given as an example. However, the invention is effective for a control system that performs another control. In addition, a multiplicity of each unit of the nuclear power plant control system described in the above Embodiment may be arbitrarily changed according to a desired degree of reliability and the like. Specifically, when detection means (corresponding to the detection units 30a to 30d) that detect a phenomenon occurring in a nuclear power plant for each of M systems are connected to a start means (corresponding to the trip control device 20) that starts processing corresponding to the phenomenon in a case where a signal that indicates an occurrence of the phenomenon is input from a predetermined number or more of signal lines out of L signal lines, a majority circuit may be provided for each of the L signal lines, and each majority circuit may output a signal that indicates an occurrence of the phenomenon to a corresponding signal line in a case where the phenomenon is detected in N or more systems out of the M systems of the detection means. However, L is an integer greater than or equal to 1, M is an integer greater than or equal to 2, and N is an integer greater than or equal to 1. 1 to 3 NUCLEAR POWER PLANT CONTROL SYSTEM 10a, 10b DETECTION UNIT 20 TRIP CONTROL DEVICE 21a, 21b INPUT UNIT 30a to 30d DETECTION UNIT 40 TRIP CONTROL DEVICE 41a to 41d INPUT UNIT 50a, 50b MAJORITY CIRCUIT 51 to 58 RELAY |
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061887498 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to an X-ray examination apparatus which includes an X-ray source, an X-ray detector, and an X-ray filter which is arranged between the X-ray source and the X-ray detector, which X-ray filter includes a plurality of filter elements having an X-ray absorptivity which can be adjusted by controlling a quantity of X-ray absorbing liquid within the individual filter elements, where individual filter elements communicate with the X-ray absorbing liquid by way of a first end. 2. Description of Related Art An X-ray examination apparatus of this kind is known from French patent application FR 2,599,886. The known X-ray examination apparatus comprises an X-ray filter for limiting the dynamic range of an X-ray image, being the interval between the extremes of the brightness values. An X-ray image is formed on the X-ray detector by arranging an object, for example a patient to be examined, between the X-ray source and the X-ray detector and by irradiating said object by means of X-rays emitted by the X-ray source. If no precautions are taken, a large dynamic range of the X-ray image may occur. On the one hand, in some parts of the object, for example lung tissue, the X-ray transmissivity will be high whereas other parts of the object, for example bone tissue, can hardly be penetrated by X-rays. If no further precautions are taken, therefore, an X-ray image with a large dynamic range is obtained whereas, for example medically relevant information in the X-ray image is contained in brightness variations in a much smaller dynamic range; because it is not very well possible to make small details of low contrast suitably visible in a rendition of such an X-ray image, such an X-ray image is not very well suited for making a diagnosis. When the X-ray image is converted, using an image intensifier pick-up chain, into a light image which is picked up by means of a video camera, the dynamic range of the light image may be much greater than the range of brightness values that can be handled by the video camera without causing disturbances in the electronic image signal. In order to limit the dynamic range of the X-ray image, the known X-ray examination apparatus includes an X-ray filter with filter elements provided with a bundle of parallel capillary tubes, each of which is connected, via a valve, to a reservoir containing an X-ray absorbing liquid which suitably wets the inner walls of the capillary tubes. In order to fill one of the capillary tubes with the X-ray absorbing liquid, the valve of the relevant capillary tube is opened after which the capillary tube is filled with the X-ray absorbing liquid by the capillary effect. Such a filled capillary tube has a high X-ray absorptivity for X-rays passing through such a filled capillary tube in a direction approximately parallel to its longitudinal direction. The valves are controlled so as to ensure that the amount of X-ray absorbing liquid in the capillary tubes is adjusted in such a manner that in parts of the X-ray beam which pass through parts of low absorptivity of the object filter elements are adjusted to a high X-ray absorptivity and that filter elements in parts of the X-ray beam which pass through parts of high absorptivity of the object, or are intercepted by a lead shutter, are adjusted to a low X-ray absorptivity. In order to change the adjustment of the filter of the known X-ray examination apparatus it is first necessary to empty filled capillary tubes. Therefore, use is made of a paramagnetic X-ray absorbing liquid which is forced out of the capillary tubes by application of a magnetic field. After all capillary tubes have been emptied, the X-ray filter is adjusted anew by deactivation of the magnetic field and by subsequently opening valves of capillary tubes which are to be filled with the X-ray absorbing liquid for the new filter adjustment so as to adjust these tubes to a high X-ray absorptivity. Consequently, it is not very well possible to change the adjustment of the known X-ray filter within a brief period of time, for example one second. Therefore, the known X-ray apparatus is not suitable for forming successive X-ray images at a high image rate while changing the adjustment of the filter between the formation of successive X-ray images. In order to control the quantity of X-ray absorbing liquid in the capillary tubes it is necessary that the period of time during which the valves are opened is accurately controlled; however, the mechanical drive of the valves, for example exhibiting inertia and play, impedes fast and accurate control of the quantity of X-ray absorbing liquid in the capillary tubes. SUMMARY OF THE INVENTION It is an object of the invention to provide an X-ray examination apparatus which includes an X-ray filter that can be adjusted more quickly than the known filter. This object is achieved by means of an X-ray examination apparatus according to the invention which is characterized in that individual filter elements communicate with an X-ray transparent liquid by way of a second end. Individual filter elements are partly filled with an X-ray absorbing liquid and their remainder is filled with an X-ray transparent liquid. In the context of the present patent application an X-ray absorbing liquid is to be understood to mean a liquid having a considerable X-ray absorptivity, for example a lead salt solution. In the context of the present application an X-ray transparent liquid is to be understood to mean a liquid which absorbs hardly any or no X-rays, for example oil. The amount of X-ray absorbing liquid in individual filter elements can be controlled hydropneumatically, i.e. on the basis of the liquid pressure in the X-ray absorbing and X-ray transparent liquids. Because only very few moving parts are required, only a very short period of time will be required so as to change the adjustment of the X-ray filter. Control of the amount of X-ray absorbing liquid on the basis of the liquid pressure also offers a faster response time in comparison with the known X-ray filter. The filter elements are preferably arranged in a matrix. Individual filter elements are arranged at intersections of respective column ducts and row ducts. Row ducts and column ducts are liquid ducts in the row direction and the column direction, respectively. The row and column directions are different directions which usually extend substantially perpendicularly to one another. It will be evident that the terms row and column can be interchanged without affecting the operation of the X-ray filter. On the basis of the difference between the liquid pressure in the relevant column duct and the relevant row duct the relevant filter element is filled or not or is filled more or less with the X-ray absorbing liquid so that the X-ray absorptivity of the relevant filter element is adjusted on the basis of the liquid pressure. By choosing a given column duct and a given row duct so as to apply a predetermined, appropriate liquid pressure thereto, the filter element at the intersection of the relevant row duct and column duct is chosen and the amount of X-ray absorbing liquid therein is thus controlled. Furthermore, it is advantageous to connect row and/or column ducts to the pressure control system by way of both ends. Consequently, only a slight pressure drop occurs in the ducts and the filter elements can be quickly and accurately adjusted to the desired X-ray absorptivity in a simple manner. It is also advantageous when the row and column ducts enclose an angle of approximately 60.degree. relative to one another. The filter elements then constitute a hexagonal pattern with a dense packing. An X-ray filter comprising a large number of filter elements per unit of surface area can be realized notably by means of cylindrical filter elements having a round cross-section. The pressure in row and/or column ducts can be controlled independently of one another by utilizing valves which are controlled by the pressure control system; in that case there will be hardly any mutual influencing between individual, for example neighboring filter elements. It is thus very well possible to form a spatial distribution of the X-ray absorption with variations over short distances by means of the X-ray filter, meaning that the X-ray filter has a high spatial resolution. A number of valves is required which amounts to approximately the square root of the number of filter elements. Thus, even if an extremely large number of filter elements is used, for example in order to achieve a high spatial resolution, the number of valves required still remains reasonable. For example, an X-ray filter comprising tens of thousands of filter elements requires only a few hundreds of valves. Preferably, the X-ray absorbing liquid is separated from the X-ray transparent liquid in the individual filter elements by pistons. The pistons counteract mixing of the X-ray transparent liquid and the X-ray absorbing liquid. Therefore, the miscibility of these liquids need not be extremely small. Furthermore, such a piston isolates the relevant filter element from the row ducts or from the column ducts when the filter element has been completely filled with one of the liquids. Due to the friction between the piston and the wall of the filter element, the adjustment of the X-ray filter is maintained and it will not be necessary to apply a liquid pressure continuously. For the design of the X-ray filter the fact is taken into account that the liquid pressure can overcome the friction between the piston and the wall of the filter element. Preferably, a coating layer is provided notably on the parts of the system which face the wall of the relevant filter element in the X-ray filter. As a result of the use of the coating layer it is achieved that no liquid can leak between the wall and the piston. Notably aluminium oxide (Al.sub.2 O.sub.3) and polyimide are suitable materials for forming such a coating layer. A high spatial resolution is achieved by means of small filter elements, preferably filter elements having a cross-section which is less than approximately 5 mm. By using an X-ray absorbing liquid and an X-ray transparent liquid which do not mix or only hardly so, mixing of the two liquids is avoided in a natural manner so that less severe requirements may be imposed as regards the sealing action of the piston. It may even be possible to dispense with the pistons. |
summary | ||
description | This application generally relates to nuclear reactor fuel assemblies and more particularly relates to a nuclear fission reactor fuel assembly adapted to permit expansion of the nuclear fuel contained therein. It is known that, in an operating nuclear fission reactor, neutrons of a known energy are absorbed by nuclides having a high atomic mass. The resulting compound nucleus separates into fission products that include two lower atomic mass fission fragments and also decay products. Nuclides known to undergo such fission by neutrons of all energies include uranium-233, uranium-235 and plutonium-239, which are fissile nuclides. For example, thermal neutrons having a kinetic energy of 0.0253 ev (electron volts) can be used to fission U-235 nuclei. Thorium-232 and uranium-238, which are fertile nuclides, undergo induced fission, with fast neutrons, which have a kinetic energy of at least 1 MeV (million electron volts). The total kinetic energy released from each fission event is about 200 MeV. This kinetic energy is eventually transformed into heat. Moreover, the fission process, which starts with an initial source of neutrons, liberates additional neutrons as well as transforms kinetic energy into heat. This results in a self-sustaining fission chain reaction that is accompanied by continued energy release. That is, for every neutron that is absorbed, more than one neutron is liberated until the fissile nuclei are depleted. This phenomenon is used in a commercial nuclear reactor to produce continuous heat that, in turn, is beneficially used to generate electricity. Fuel assembly expansion due to the aforementioned heat generation and also due to fission product release can occur in such processes. In this regard, fuel assemblies may undergo differential expansion, fuel rod creep that can increase incidence of fuel rod cladding rupture, fission gas pressure build-up, and swelling during reactor operation. This may increase the incidence of fuel pellet cracking and/or fuel rod bowing. Fuel pellet cracking may lead to fission gas release and cause higher than normal radiation levels. Fuel rod bowing may in turn lead to obstruction of coolant flow channels. Safety margins incorporated into the reactor design and precise quality control during manufacture can reduce these incidences or the system design can adapt systems to operate with such incidences. In one approach to deading with fuel assembly expansion due to heat generation and fission gas release, U.S. Pat. No. 3,028,330 issued Apr. 3, 1962 in the name of Clarence I. Justheim, et al. and titled “Nuclear Fuel Elements Having An Autogenous Matrix And Method Of Making The Same” discloses a cellular carbonaceous matrix. The cells of the cellular matrix can contain fragments of a fissile material, which may be fertile isotopes of uranium enriched with fissionable isotopes. According to this patent, the cells are ordinarily of such size relative to the fission fragments as to allow for increase in volume of the latter resulting from thermal cycling and radiation damage. Although this patent discloses a cellular matrix having cells that allow for increase in volume of fission fragments, this patent does not appear to disclose a nuclear fission reactor fuel assembly, which is adapted to permit expansion of the nuclear fuel contained in the fuel assembly. Another approach, disclosed in U.S. Pat. No. 3,184,392 issued May 18, 1965 in the name of Leslie Reginald Blake, et al. and titled “Fast Nuclear Reactor Fuel Elements” describes a nuclear reactor fuel element that comprises a body of porous, closed-cell, fissile nuclear fuel which by virtue of the porosity is of dispersed structure and which is enclosed within a cylindrical protective sheath. The fuel provides interstitial voids and the fuel element is only partially filled to leave a void above the fuel. According to this patent, the protective sheath is capable of withstanding an internal pressure of at least 10,000 psi at a temperature of 600° C. and the void above the fuel serves as expansion space for the fuel and also space for accommodation of fission products. Although this patent discloses a nuclear reactor fuel element that comprises a body of porous, closed-cell, fissile nuclear fuel, this patent does not appear to disclose a nuclear fission reactor fuel assembly, which is adapted to permit expansion of the nuclear fuel contained in the fuel assembly, as disclosed and claimed herein. According to an aspect of this disclosure, there is provided a nuclear fission reactor fuel assembly, comprising an enclosure adapted to sealingly enclose a nuclear fuel foam defining a plurality of interconnected open-cell voids. According to another aspect of this disclosure, there is provided a nuclear fission reactor fuel assembly, comprising an enclosure adapted to sealingly enclose a fertile nuclear fuel foam defining a plurality of closed-cell voids. According to yet another aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure adapted to sealingly enclose a nuclear fuel foam capable of generating heat, the nuclear fuel foam defining a plurality of interconnected open-cell voids; and a heat absorber associated with the enclosure and adapted to be in heat transfer communication with the nuclear fuel foam for absorbing the heat generated by the nuclear fuel foam. According to a further aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure adapted to sealingly enclose a fertile nuclear fuel foam capable of generating heat, the nuclear fuel foam defining a plurality of closed-cell voids; and a heat absorber associated with the enclosure and adapted to be in heat transfer communication with the nuclear fuel foam for absorbing the heat generated by the nuclear fuel foam. According to another aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure adapted to sealingly enclose a nuclear fuel foam capable of generating heat, the nuclear fuel foam defining a plurality of interconnected open-cell voids; and a heat absorber conduit extending through the nuclear fuel foam, the heat absorber conduit capable of carrying a cooling fluid therealong in heat transfer communication with the nuclear fuel foam for absorbing the heat generated by the nuclear fuel foam. According to yet another aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure adapted to sealingly enclose a nuclear fuel foam capable of generating heat, the nuclear fuel foam defining a plurality of closed-cell voids; and a heat absorber conduit extending through the nuclear fuel foam, the heat absorber conduit capable of carrying a cooling fluid therealong in heat transfer communication with the nuclear fuel foam for absorbing the heat generated by the nuclear fuel foam. According to a further aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure; a nuclear fuel foam capable of generating heat, the nuclear fuel foam being sealingly disposed in the enclosure, the nuclear fuel foam defining a plurality of interconnected open-cell voids; and a heat absorber disposed in heat transfer communication with the nuclear fuel foam for absorbing the heat generated by the nuclear fuel foam. According to still another aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure; a fertile nuclear fuel foam capable of generating heat, the nuclear fuel foam sealingly disposed in the enclosure, the nuclear fuel foam defining a plurality of closed-cell voids; and a heat absorber disposed in heat transfer communication with the nuclear fuel foam for absorbing the heat generated by the nuclear fuel foam. According to another aspect of the disclosure, there is provided a nuclear fission reactor fuel assembly, comprising: an enclosure adapted to sealingly enclose a porous uncoated nuclear fuel material defining a plurality of interconnected open-cell voids. According to yet another aspect of the disclosure, a method includes, but is not limited to, making a nuclear fission reactor fuel assembly, comprising the step of providing an enclosure to sealingly enclose a nuclear fuel foam defining a plurality of interconnected open-cell voids. According to yet another aspect of the disclosure, a method includes, but is not limited to, making a nuclear fission reactor fuel assembly, comprising the step of providing an enclosure to sealingly enclose a fertile nuclear fuel foam defining a plurality of closed-cell voids. According to a further aspect of the disclosure, a method includes, but is not limited to, operating a nuclear fission reactor fuel assembly, comprising the step of disposing an enclosure in a nuclear reactor vessel, the enclosure sealingly enclosing a nuclear fuel foam defining a plurality of interconnected open-cell voids. According to yet another aspect of the disclosure, a method includes, but is not limited to, operating a nuclear fission reactor fuel assembly, comprising the step of disposing an enclosure in a nuclear reactor vessel, the enclosure sealingly enclosing a fertile nuclear fuel foam defining a plurality of closed-cell voids. In addition to the foregoing, other method aspects are described in the claims, drawings, and text forming a part of the present disclosure. A feature of the present disclosure is the provision of an enclosure adapted to sealingly enclose a nuclear fuel foam defining a plurality of interconnected open-cell voids. Another feature of the present disclosure is the provision of an enclosure adapted to sealingly enclose a fertile nuclear fuel foam defining a plurality of closed-cell voids. In addition to the foregoing, various other method and/or device aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented herein. In addition, the present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Moreover, the herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. Therefore, turning to FIG. 1, there is shown a nuclear fission reactor, generally referred to as 10, for producing heat due to fission of a fissile nuclide, such as uranium-235, uranium-233, or plutonium-239. Reactor 10 may be a “traveling wave” reactor. In this regard, a traveling wave reactor includes a reactor core. A nuclear fission igniter in the reactor core initiates a fission deflagration wave burnfront. After the nuclear fuel in the core is ignited by the fission igniter, the fission deflagration wave burnfront is initiated and propagates throughout the nuclear fuel. In one embodiment, during this fission process, reactor coolant loops transfer heat from the reactor core to heat exchangers for producing steam. The steam is transferred to a turbine-generator for generating electricity. Such a traveling wave reactor is disclosed in more detail in co-pending U.S. patent application Ser. No. 11/605,943 filed Nov. 28, 2006 in the names of Roderick A. Hyde, et al. and titled “Automated Nuclear Power Reactor For Long-Term Operation”, which application is assigned to the assignee of the present application, the entire disclosure of which is hereby incorporated by reference. Still referring to FIG. 1, reactor 10 comprises a vessel 20, such as a pressure vessel or containment vessel, for preventing leakage of radioactive particles, gasses or liquids from reactor 10 to the surrounding environment. Vessel 20 may be steel, concrete or other material of suitable size and thickness to reduce risk of such radiation leakage and to support required pressure loads. Although only one vessel 20 is shown, there may be additional containment vessels, one enveloping the other, for added safety. Vessel 20 defines a well 30 therein in which is disposed one or more nuclear fission reactor fuel assemblies 40, as described in more detail hereinbelow. As best seen in FIGS. 2, 3 and 4, each nuclear fission reactor fuel assembly 40 comprises a generally cylindrical enclosure 50 having enclosure walls 60 for sealingly enclosing a nuclear fuel foam 70 therein. Foam 70 defines a plurality of closed-cell voids 80 spatially distributed within it. As used herein, the terminology “closed-cell voids” means that each void 80 is separated from and typically not interconnected to its neighboring void 80, such that substantial amounts of gas, liquid, or fluid do not directly travel between voids 80. As seen in FIG. 5, foam or porous material 70 may alternatively define a plurality of interconnected open-cell voids 90 spatially distributed within it. As used herein, the terminology “open-cell voids” means that each void is typically connected to one or more of its neighbors, permitting gas, liquid, or fluid to directly travel between voids 90. The open-cell voids may be defined by a foam fuel material having a web-like or honeycomb structure. The open-cell voids may be defined by a porous fuel material having a fibrous or rod-like structure, or a porous fuel material formed by an interconnected collection of fuel particles (such as sintered beads or packed spheres). Also, the open-cell voids may be defined by fuel material having a mixture of foam or porous characteristics. Foam or porous material 70 may comprise a fissile nuclear fuel, such as uranium-233, uranium-235 and/or plutonium-239. Alternatively, foam 70 may comprise a fertile nuclear fuel, such as thorium-232 and/or uranium-238. A further alternative is that foam or porous material 70 may comprise a predetermined mixture of fissile and fertile nuclear fuel. It will be appreciated by a person of ordinary skill in the art that fuel assembly 40 may be disposed in a thermal neutron reactor, a fast neutron reactor, a neutron breeder reactor, a fast neutron breeder reactor or the previously mentioned traveling wave reactor. Thus, fuel assembly 40 is versatile enough to be beneficially used in various nuclear reactor designs. With reference to FIGS. 2, 3, 4 and 5, it will be understood that a purpose of each void 80 and 90 is to provide a shrinkable volume that is adapted to accommodate or permit expansion of foam or porous material 70 due to thermal expansion and fission product gas release during operation of reactor 10. Overall, void volume of foam or porous material 70 may be approximately 20% to approximately 97% to permit the expansion, although in come cases percentages outside of this range may be producible. Accommodating expansion of foam or porous material 70 in this manner reduces pressure on enclosure walls 60 because foam or porous material 70 will expand toward or even into voids 80 or 90 rather than expand against walls 60. Therefore, this structure is typically configured such that foam or porous material 70 expands inwardly toward voids 80 or 90 rather than outwardly against enclosure walls 60 to exert pressure on enclosure walls 60. Reducing pressure on enclosure walls 60 in turn reduces risk of enclosure 50 swelling and enclosure walls 60 cracking, both of which might otherwise lead to release of fission products. With reference to FIG. 5, it will be understood that a purpose of the interconnected open-cell voids 90 is to provide a path to facilitate transport of volatile fission products generated by the nuclear fuel foam or porous material 70. Such fission products may be isotopes of iodine, bromine, cesium, potassium, rubidium, strontium, xenon, krypton, barium or other gaseous or volatile materials. Such a transport path may provide a vehicle to remove a portion of fission products from neutronically active regions of the nuclear fission reactor fuel assembly 40. Such removal may reduce neutron absorption by fission products. Referring to FIGS. 2, 3, 4, and 5, foam or porous material 70 may substantially comprise a metal, such as uranium, thorium, plutonium, or alloys thereof. Alternatively, foam or porous material 70 may substantially comprise a carbide, such as uranium carbide (UC or UCx) or thorium carbide (ThC2 or ThCx). The uranium carbide or thorium carbide may be sputtered into a matrix of niobium carbide (NbC) and zirconium carbide (ZrC). A potential benefit of using niobium carbide and zirconium carbide is that they form a refractory structural substrate for the uranium carbide or thorium carbide. Foam or porous material 70 may also substantially comprise an oxide, such as uranium dioxide (UO2); thorium dioxide (ThO2), which is also referred to as thorium oxide; or uranium oxide (U3O8). On the other hand, foam or porous material 70 may be a nitride, such as uranium nitride (U2N3) or thorium nitride (ThN). Moreover, the discussion hereinabove related to foam or porous material 70 as being uncoated. If desired, foam or porous material 70 may be coated with a suitable material. Referring to FIGS. 6 and 7, foam or porous material 70 may be coated with a coating layer 100, which may comprise carbon, zirconium carbide or the like. Processes to achieve the desired coating may be electro-plating, electroless deposition, vapor deposition, ion deposition or any other suitable process. The coating foam or porous material 70 may provide a barrier to escape of fission products from foam or porous material 70 into voids 80 or 90. Such fission products may be isotopes of iodine, bromine, cesium, potassium, rubidium, strontium, xenon, krypton, barium or other gaseous or volatile materials. Coating foam or porous material 70 may also or alternatively provide structural support to the foam or porous material 70. Returning to FIG. 2, a heat absorber, generally referred to as 110, is associated with enclosure 50 and is adapted to be in heat transfer communication with foam or porous material 70 for absorbing fission heat generated by foam or porous material 70. By way of example only, and not by way of limitation, heat absorber 110 may comprise a plurality of generally cylindrical parallel conduits or pipes 120 extending through foam or porous material 70. Each pipe 120 has a pipe wall 130 defining a flow channel 140 for reasons described presently. Pipes 120 may be fabricated from refractory metals or alloys such as Niobium (Nb), Tantalum (Ta), tungsten (W) or the like. The reactor coolant pipes 120 may be made from other materials such as aluminum (Al), steel or other ferrous or non-iron group alloys or titanium or zirconium-based alloys, or from other suitable metals and alloys. A coolant, such as pressurized gas (not shown), flows along flow channel 140 for absorbing heat from foam or porous material 70 by means of heat conduction through pipe wall 130. The reactor coolant may be selected from several pressurized inert gases, such as helium, neon, argon, krypton, xenon, or mixtures thereof. Alternatively, the coolant may be water, or gaseous or superfluidic carbon dioxide, or liquid metals, such as sodium or lead, or liquid metal alloys, such as lead-bismuth (Pb—Bi), or organic coolants, such as polyphenyls, or fluorocarbons. Alternatively, the coolant may be a phase-changing composition, such as water, potassium (K) or sodium (Na). On the other hand, the heat absorber 110 may be a thermoelectric material, such as bismuth telluride (Bi2Te3); lead telluride (PbTe); or zinc antimonide(Zn4Sb3). It is appreciated by those of ordinary skill in the art that heat absorber 110 or pipes 120 need not be parallel, as shown; rather, heat absorber 110 or pipes 120 may be set at criss-cross angles with respect to each other, if desired. Turning now to FIGS. 8 and 9, an alternative embodiment of fuel assembly 40 is there shown. In this alternative embodiment, fuel assembly 40 comprises a generally spherical enclosure 150, rather than the previously mentioned generally cylindrical enclosure 50. The spherical enclosure 150 may reduce the amount of cladding material required. The spherical enclosure 150 may also help shape fuel profiles. Referring to FIGS. 10 and 11, yet another alternative embodiment of fuel assembly 40 is there shown. In this alternative embodiment, fuel assembly 40 comprises a generally disk-shaped enclosure 160. A potential benefit to using the disk-shaped enclosure 160 is its utility in shaping fuel profiles. Referring to FIGS. 12 and 13, another alternative embodiment of fuel assembly 40 is there shown. In this alternative embodiment, fuel assembly 40 comprises a generally hemi-spherical enclosure 155. The_hemi-spherical enclosure 155 may increase fuel assembly packing densities in well 30. As with the spherical profile, the hemi-spherical enclosure 155 may aid in shaping fuel profiles. Referring to FIGS. 14 and 15, still another embodiment of fuel assembly 40 is there shown. In this alternative embodiment, fuel assembly 40 comprises a polygonal-shaped (in transverse cross-section) enclosure 165. In this regard, enclosure 165 may have a hexagon shape in transverse cross section. A potential benefit to using the hexagonally shaped cross section of enclosure 165 is that more fuel assemblies 40 may provide relatively high packing factors in some configurations and increase the number of fuel assemblies packed into well 30 over some other geometric shapes for the fuel assembly. As with the previous embodiments, the hexagonally shaped cross section of enclosure 165 may assist in shaping fuel profiles. Referring to FIGS. 16 and 17, yet another alternative embodiment of fuel assembly 40 is there shown. In this alternative embodiment, fuel assembly 40 comprises a parallelepiped-shaped enclosure 170. The parallelepiped-shaped enclosure 170 may also provide relatively high packing density in well 30. As with the previous embodiments, the parallelepiped-shaped enclosure 170 may assist in shaping fuel profiles. Referring to FIG. 18, foam or porous material 70 may include one or more fuel pellets 180 embedded therein. In one embodiment, fuel pellet 180 may serve as an initial source of reactivity for starting the previously mentioned fission chain reaction. In another embodiment, fuel pellet 180 may serve as a higher density fuel component to increase the effective density of the nuclear fuel material. As reactor 10 is operated, foam or porous material 70 will tend to expand. This can occur because during operation of reactor 10, foam or porous material 70 will undergo thermal expansion due to fission heat generated by foam or porous material 70 during the fission process. Fission gasses will also be produced due to the fission process. These two phenomena will tend to expand foam or porous material 70, which in turn may tend to put pressure on enclosure wall 60. Such pressure may increase the risk of a breach of enclosure wall 60 and subsequent release of fission products from fuel assembly 40. The foam or porous material 70 disclosed herein addresses this effect by providing shrinkable voids 80 and 90. In other words, voids 80 and 90 can accommodate or permit expansion of foam or porous material 70 by being reduced in volume as foam or porous material 70 expands toward voids 80 and 90. In this manner, the potential pressure increase on wall 60 is reduced and risk of fission product release from fuel assembly 40 is likewise reduced. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken as limiting. Moreover, those skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Therefore, what is provided is a nuclear fission reactor fuel assembly, as described and claimed herein, which is adapted to permit expansion of the nuclear fuel contained in the fuel assembly. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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abstract | Methods of fabricating a photo mask are provided. The method includes collecting sample data, setting a preliminary mask layout, performing an optical proximity correction using the sample data and a preliminary mask layout to obtain an optimized preliminary mask layout, verifying the optimized preliminary mask layout to obtain a final mask layout, and fabricating the photo mask using the final mask layout. Verification of the optimized preliminary mask layout includes operating a verification simulator using the sample data and the optimized preliminary mask layout as input data to obtain verification image data. The verification image data includes a plurality of contours of a pattern at different vertical positions. |
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claims | 1. A method of detecting and forecasting resource bottlenecks of a computer system comprising:monitoring with successive measurements a utilization parameter of a system resource;computing a change parameter by comparing the differences between successive measurements of the utilization parameter;comparing the change parameter to a threshold change parameter;reporting a resource bottleneck if the change parameter exceeds the threshold change parameter; anddetecting false bottleneck alarms and modifying the threshold change parameter based on the false bottleneck alarms to decrease a sensitivity of the method. 2. The method of claim 1 further comprising detecting bottlenecks that are not reported resource bottlenecks and modifying the threshold change parameter based on detecting bottlenecks that are not reported resource bottlenecks to increase a sensitivity of the method. 3. The method of claim 1 wherein reporting the resource bottleneck if the change parameter exceeds the threshold change parameter further comprises delaying reporting the resource bottleneck until the change parameter exceeds the threshold change parameter on at least one successive measurements. 4. The method of claim 1 wherein the utilization parameter includes an average utilization of the system resource for a time period and wherein computing a change parameter by comparing the differences between successive measurements of the utilization parameter comprises subtracting successive measurements of the utilization parameter, and wherein the utilization parameter is distributed in sequentially consecutive utilization classes of increasing utilization, the average utilization for each time period being established for each utilization class, and wherein computing the change parameter comprises comparing the difference between average utilization for consecutive classes at least at two different time periods. 5. The method of claim 1 wherein the utilization parameter includes a standard deviation of the utilization of the system resource for a time period and wherein computing a change parameter by comparing the differences between successive measurements of the utilization parameter comprises determining if the utilization of the system is increasing and the standard deviation of the utilization of the system resource is decreasing based on the successive measurements. 6. The method of claim 1 wherein the utilization parameter is the median load of the utilization of the system resource for a time period and wherein computing a change parameter by comparing the differences between successive measurements of the utilization parameter comprises determining if the median load is less than the utilization of the system and then greater than the utilization of the system on a successive measurement. 7. A computer program product comprising:a computer useable medium having computer readable code means embodied thereon for causing a computer to execute a method for detecting and forecasting resource bottlenecks of a computer system, the computer readable code means in the computer program product including:computer readable program code means for causing a computer to monitor with successive measurements a utilization parameter of a system resource;computer readable program code means for causing a computer to compute a change parameter by comparing the differences between successive measurements of the utilization parameter;computer readable program code means for causing a computer to compare the change parameter to a threshold change parameter;computer readable program code means for causing a computer to report a resource bottleneck if the change parameter exceeds the threshold change parameter; andcomputer readable program code means for causing a computer to detect false bottleneck alarms and to modify the threshold change parameter based on the false bottleneck alarms to decrease a sensitivity. 8. The computer program product of claim 7 further comprising computer readable program code means for causing a computer to detect bottlenecks that are not reported resource bottlenecks and to modify the threshold change parameter based on detected bottlenecks that are not reported resource bottlenecks to increase a sensitivity. 9. The computer program product of claim 7 wherein the resource bottleneck is not reported until the change parameter exceeds the threshold change parameter on at least one successive measurement. 10. The computer program product of claim 7 wherein the utilization parameter is the average utilization of the system resource for a time period. 11. The computer program product of claim 7 wherein the utilization parameter is the standard deviation of the utilization of the system resource for a time period. 12. The computer program product of claim 7 wherein the utilization parameter is the median load of the utilization of the system resource for a time period. 13. A data processing system comprising:a processor; anda program code executed on the processor for detecting and forecasting resource bottlenecks, the program code including code for:monitoring with successive measurements a utilization parameter of a system resource;computing a change parameter by comparing the differences between successive measurements of the utilization parameter;comparing the change parameter to a threshold change parameter; andpredicting a resource bottleneck if the change parameter exceeds the threshold change parameter;wherein the program code further includes code for detecting false bottleneck alarms and modifying the threshold change parameter based on the false bottleneck alarms to decrease a sensitivity. 14. The data processing system of claim 13 wherein the program code further includes code for determining a corrective action to avoid the resource bottleneck. 15. The data processing system of claim 14 wherein the data processing system is a server within a LAN network and the utilization parameter is a percentage of CPU utilization. 16. The data processing system of claim 13 wherein the program code further includes code for detecting bottlenecks that are not reported resource bottlenecks and modifying the threshold change parameter based on detecting bottlenecks that are not reported resource bottlenecks to increase a sensitivity of the method. 17. The data processing system of claim 13 wherein the program code further includes code for reporting the resource bottleneck if the change parameter exceeds the threshold change parameter on at least one successive measurement. |
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048184683 | claims | 1. A process for preparing .sup.123 I which comprises: irradiating with a proton beam in an irradiation zone a liquid sample of XI as a target material, wherein X is alkali metal or I, the protons in said beam having an energy in the range of 60-70 MeV at a power level of at least about 5 .mu.a, wherein the thickness of the XI sample is sufficient to reduce the energy of said particles by from about 15-25 MeV and said beam irradiating a substantial portion of said liquid sample in said zone; continuously passing a stream of helium onto the surface of said liquid sample in said irradiation zone whereby .sup.123 Xe produced by said irradiation is entrained with said helium and carried from the irradiation zone to a condensing zone; cooling said helium stream in said condensing zone returning XI to said irradiation zone and transferring said helium stream to a collection zone; condensing .sup.123 Xe by cooling said helium stream in said collection zone to a temperature below the condensation temperature of .sup.123 Xe, and collecting .sup.123 Xe in said collection zone; and allowing .sup.123 Xe to decay to .sup.123 I. irradiating with a scanning proton beam in an irradiation zone a liquid sample of XI as a target material, wherein X is sodium or I, the protons in said beam having an energy in the range of 60-70 MeV at a power level of from about 10 .mu.a to 20 .mu.a, wherein the thickness of the XI sample is sufficient to reduce the energy of said particles by from about 15-25 MeV, while maintaining the temperature at or about the melting point of XI, but not to exceed 100.degree. C. above the melting point of XI and said beam irradiating a substantial portion of said liquid sample in said zone; continuously passing onto the surface of said target material in said irradiation zone a helium stream containing up to 0.2 volume percent xenon, whereby .sup.123 Xe produced by said irradiation is entrained with said helium and carried from said irradiation zone to a condensing zone; cooling said helium stream in said condensing zone and returning condensed XI to said irradiation zone; removing any entrained I.sub.2 from said helium stream exiting from said condensing zone and transferring the substantially I.sub.2 free helium stream to a collection zone; condensing .sup.123 Xe by cooling said helium stream in said collection zone to a temperature below the condensation temperature of .sup.123 Xe and collecting .sup.123 Xe in said collection zone; and transferring said condensed .sup.123 Xe in vacuo by evaporation and condensation from said collection zone to a decay zone where .sup.123 Xe decays to .sup.123 I to provide high purity .sup.123 I. 2. A method according to claim 1, wherein said condensed .sup.123 Xe is transferred in vacuo by evaporation and condensation to a second vessel at a pressure below about 50.mu. mercury to enhance the purity of the .sup.123 I obtained from decay of .sup.123 Xe. 3. A method according to claim 1, wherein said helium stream has from about 0-0.2 volume percent xenon and entrained I.sub.2 is removed from said helium stream exiting from said condensing zone. 4. A method according to claim 1, wherein XI is sodium iodide and the temperature of said target material is in the range of about 650.degree.-720.degree. C. 5. A method according to claim 1, wherein XI is I.sub.2 and the temperature of said target material is in the range of about 100.degree.-130.degree. C. 6. A method according to claim 1, wherein said helium stream is passed through said target material at a rate of about 10-60 ml per minute. 7. A method according to claim 1, wherein said helium stream is heated by said liquid sample prior to passing onto the surface of said liquid sample. 8. A method according to claim 1, wherein XI is placed in tandem in a second target zone behind said target zone having a thickness sufficient to reduce the energy of the protons' radiation exiting from said first zone by 15-20 MeV; and isolating .sup.125 Xe produced by said radiation exiting from said first zone. 9. A method for preparing .sup.123 I which comprises: 10. A method according to claim 9, where XI is NaI. 11. A method according to claim 9, where XI is I.sub.2. 12. A method according to claim 9, wherein said helium stream is heated by said target material prior to passing onto the surface of said target material. |
042648246 | description | DESCRIPTION Collimator 10 has spaced, radiation absorbing sheets 12, and may be of the same general construction described in said patent and application. However, sheets 12, rather than being parallel, lie along planes 14 the continuations of which intersect in a common focal axis 16. In a typical preferred embodiment the collimator would have 100 sheets 12, equally angularly spaced to give a total field of view of 10.degree. about axis 16. Sheets 12 extend along axis 16 sufficiently to give about the same 10.degree. field of view about axis 18 perpendicular to axis 16. A separate detector strip 20, parallel to axis 16, is mounted between the converging ends of each pair of sheets 12. The detectors are of the sort described in said patent. The collimator and detectors are mounted (using appropriate and conventional hardware, not shown) for oscillation through 180.degree. about an axis 22 perpendicular to axes 16 and 18, under the control of drive unit 24. Conventional signal processing circuitry 26 is connected to detectors 20 to initially process the information sensed during operation of the device. Computer 28 is provided for data reduction. Stepping motor drive unit 24 provides clock inputs to the computer. The extent of sheets 12 along planes 14 is such that each central viewing plane 30, bisecting a collimator slit 31, is parallel to the "boundary" viewing plane 32 (which runs between the forward edge 34 of one sheet 12 and the rear edge 36 of the other sheet defining the slit) of an adajacent slit 31. OPERATION The instrument is set up by pointing axis 22 at the estimated position of the radiation source to be imaged, e.g., a cellestial body, to bring the source into the overall field of view of the collimator. Drive 24 is turned on and collimator 10 is rotated, in steps, 180.degree. about axis 22. After each step detectors 20 are turned on and readings are taken. For a given angular position of the collimator about axis 22, radiation from a given source within the field of view will be received primarily through one slit (by its associated detector 20), though there will be some reception through neighboring slits. In other words, the effect of making sheets 12 non-parallel is that, for a given angular position of the collimator, the slits have different transmissivities of radiation from a given source. Rotation of the collimator serves to vary those transmissivities with time. In general, circuitry 26 will note the response of each detector 20 for each angular position of the collimator about axis 22. Data defining these responses, and the angular positions about axis 22 at which they occur, is fed to computer 28. Using a data reduction procedure substantially identical to that described in said patent, angular coordinates (rather than Cartesian coordinates as in the near field case) of each radiation source within the field of view are computed. Embodiments of the invention using diverging sheets are also useful in near field imaging. Other embodiments are within the following claims. For example, instead of stepping the collimator, it might be rotated continuously, with the time of arrival of each photon, and the slit through which it arrived, being recorded. Further, the slit-to-slit difference in transmissivity might be achieved other than by angling the sheets, and the variation of transmissivity with time other than by rotating the collimator. |
description | In accordance with the invention, an X-ray system with symmetric collimation as described above may be readily modified for chest imaging, while minimizing the impact of the modification on the overall system architecture, and without losing important benefits and advantages of the system. Such modification is usefully understood by first referring to FIG. 1, which shows an X-ray tube 10 disposed to project a beam of X-rays through a collimator 12 adjacent thereto. The projected beam is incident upon the detector plane 14a of a digital solid-state X-ray detector 14, within an active imaging area or AIA. Collimator 12 is provided with internal collimator blades 16 and 18, which are movable toward or away from each other to respectively decrease or increase the vertical dimension of the AIA. FIG. 1 shows collimator blades 16 and 18 adjusted to provide an X-ray beam 20 bounded by an upper ray 20a and a lower ray 20b, which are directed to the upper and lower edges, respectively, of detector 14. Thus, collimator 12 in FIG. 1 provides a full field view, that is, the vertical dimension of beam 20 and the vertical dimension of detector plane 14a of detector 14 coincide at their intersection. In FIG. 1, internal collimator blades 16 and 18 provide symmetric collimation in that for any adjustment thereof, the X-ray beam projected therethrough will be symmetric about the focal point center line 22 of the beam. Referring to FIG. 2, there is shown the arrangement depicted in FIG. 1, modified in accordance with the invention by joining a fixed collimator blade 24 to the collimator 12. More specifically, collimator blade 24 is fixably joined to collimator 12 so that it lies in the path of the upper portion of an X-ray beam projected out from collimator 12. Thus, upper ray 20a of beam 20 is blocked by collimator blade 24. As shown by FIG. 2, the uppermost portion of any beam which may be projected by collimator 12 is bounded by ray 26a. As described hereinafter in connection with FIG. 4, detector 14 is vertically adjustable with respect to X-ray tube 10 and collimator 12. Accordingly, FIG. 2 further shows detector 14 vertically adjusted so that ray 26a of a projected beam intersects the upper edge of detector 14. To provide a full field of view on the detector plane 14a, so that the entire area of the detector plane may be used to acquire an image, collimator blades 16 and 18 are selectively moved apart, with respect to their positions as shown in FIG. 1, to provide an X-ray beam 28. Beam 28, which is wider than beam 20, has a lower ray 28b which intersects the lower edge of detector 14. However, upper ray 28a of beam 28 is blocked by the fixed collimator blade 24. Referring to FIG. 3, there is shown collimator blades 16 and 18 moved toward one another, with respect to their positions shown in FIG. 2, to project an X-ray beam 30 having an upper ray 30a and a lower ray 30b. Upper ray 30a is blocked by fixed collimator blade 24, and ray 26a continues to intersect the upper edge of detector 14. However, ray 30b intersects the detector plane 14a at a distance Xclose above the lower edge thereof. Thus, a lower region 14b of detector 14 does not receive X-rays of beam 30. Moreover, the dimension xclose of lower region 14b may be readily changed by adjustment of collimator blades 16 and 18, up to a limit described hereinafter, while the remainder of the detector continues to receive X-radiation. Accordingly, the collimator arrangement shown in FIGS. 2 and 3 is operable to asymmetrically collimate an X-ray beam projected therethrough. Such arrangement may be readily employed for chest imaging as described above. A patient would be positioned directly in front of detector 14, with his or her chin near the top of the detector. Collimator blades 16 and 18 would then be adjusted to bring lower region 14b of detector 14 and the lower abdominal region of the patient, which is to be excluded from X-ray exposure, into coincident relationship. Thus, the arrangement of FIGS. 2 and 3 allows symmetric X-ray collimator 12 to be used in an application requiring asymmetric collimation, while at the same time maintaining image quality. In addition, the effective radiation received by a patient can be reduced without the use of cumbersome shields or aprons of the type described above. Moreover, in products such as the Revolution XQ/i, referred to above, a user will be able to position the patient in a customary manner utilizing the auto tracking tube designed for overall customer productivity. Collimator 12 is also provided with a pair of internal collimator blades (not shown) which are used to adjust the horizontal dimension of the AIA. Though available, such collimator blades are generally not used in connection with the present embodiment of the invention. Referring to FIG. 4, there is shown tube 10 and collimator 12 mounted for selected vertical positioning along a support column 32. Detector 14 is likewise shown to be mounted for vertical positioning along a support column 34. There is further shown collimator 12 adjusted as previously described in connection with FIG. 2, so that the X-ray system of FIG. 4 is set up for full field of view operation. That is, the AIA coincides with the entire detector plane 14a of detector 14. This is illustrated by FIG. 5. which also depicts a region 24a directly above detector 14. Region 24a represents an area from which X-rays projected by tube 10 are blocked by the action of fixed blade 24. Referring further to FIG. 4, there is shown the center line 36 of detector 14 offset from focal spot center line 22 of the X-ray tube 10 by an amount Xoffset, due to the effect of fixed collimator blade 24. Referring to FIG. 6, there is shown collimator 12 adjusted as previously described in connection with FIG. 3. Thus, the AIA on detector 14 is reduced by lower region 14b of length Xclose, as described above and as illustrated by FIG. 7. As collimator blades 16 and 18 are moved closer together, the X-ray field of view on detector 14 will continue to close from the bottom of the detector upward, until Xclose=2xoffset. At this point the field of view will converge together from the top and bottom of detector 14 at an equal rate, until the collimator blades 16 and 18 close together at focal spot center line 22. Obviously, many other modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that within the scope of the disclosed concept, the invention may be practiced otherwise than as has been specifically described. |
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description | This invention was made with Government support under contract number DE-FC07-07ID14778 awarded by the United States Department of Energy. The Government has certain rights in the invention. 1. Field of the Invention In general, the invention relates to the monitoring of thermal neutron flux within a nuclear reactor, and in particular to an optical gamma thermometer for use in a monitoring string having a local power range monitor in which the measured temperature from the optical gamma thermometer, in conjunction with a steady-state heat balance, is used to calibrate the local power range monitor during its in-service lifetime. 2. Description of the Related Art In the nuclear reaction interior of conventional boiling water reactors (BWR), it is possible to monitor the state of the reaction by either the measurement of thermal neutron flux, or alternatively gamma ray flux. Thermal neutron flux is the preferred measurement. As it is directly proportional to power and provides for a prompt (instantaneous) signal from a fission chamber. The alternative measurement of gamma radiation does not have the required prompt response necessary for reactor safety requirements. Consequently, gamma radiation as measured by gamma thermometers is not used to measure and immediately control the state of a reaction in boiling water nuclear reactors. Boiling water reactors have their thermal neutron flux monitored by local power range monitors, otherwise known as a local power range monitoring (LPRM) system. These local power range monitors include a cathode having fissionable material coated thereon. The fissionable material is usually a mixture of U235 and U234. The U235 is to provide a signal proportional to neutron flux and the U234 to lengthen the life of the detector. The thermal neutrons interact with the U235 and cause fission fragments to ionize an inert gas environment, typically argon, in the interior of the conventional local power range monitor. There results an electric charge flow between the anode and cathode with the resultant DC current. The amperage of the DC current indicates on a substantial real time basis the thermal neutron flux within the reactor core. The boiling water reactor local power range monitors are inserted to the core of the reactor in strings. Each string extends vertically and typically has four spaced apart local power range monitors. Each detector is electrically connected for reading the thermal neutron flux in real time and for outputting the state of the reaction within the reactor. It is to be understood that a large reactor can have on the order of 30 to 70 such vertical strings with a total of about 120 to 280 local power range monitors. Such local power range monitors use finite amounts of U235 during their in-service life. Consequently, the sensitivity changes with exposure and they must be periodically calibrated. Calibration is presently accomplished by using traversing in-core probes (TIPs). These traversing in-core probes are typically withdrawn from the reactor, as the traversing in-core probes are of the same basic construction as the local power range monitors and thus change their sensitivity with in-service life due to uranium 235 burnup. In operation, the traversing in-core probes are typically calibrated. Such calibration includes inserting about five such probes separately to a common portion of a boiling water reactor. The boiling water reactor is operated at steady state and made the subject of an energy balance of a type well known in the art. The insertion of the traversing in-core probes occurs by placing the probes at an end of a semi-rigid cable and effecting the insertion within a tube system. Once a full core scan has occurred during steady state operation, a heat balance is utilized in combination with the readings of the traversing in-core probes to calibrate the local power range monitors. In-core probes travel through the reactor in a specially designed tube system. This tube system constitutes through containment conduits into the interior of the reactor vessel. Into these conduits are placed semirigid cables which cables have the TIPs on the distal end thereof. The TIPs are driven into the drive tube system from large drive mechanisms and the entire system is controlled from an electronic drive control unit. The cables pass through so-called “shear valves” which valves can shear the cable and seal the conduit to prevent through the tube system leaks, which leaks may well be substantial before the cable and probes could be withdrawn. The cables further pass through stop valves admitting the traversing in-core probes to the interior of the vessel containment. Finally, the cables reach so-called indexers, and then to the interior of the reactor vessel. These indexers provide a mechanical system for routing each of the TIPs to pass adjacent the site of an assigned segment of the 170 some odd local power range monitors in a large boiling water nuclear reactor. It is normal for an indexer to include 10 alternative paths for a single traversing in-core probe to follow during a calibration procedure. Needless to say, this system is elaborate and complex. Calibration of each local power range monitor is a function of the probe measurement of the local thermal neutron flux as well as a function of the position of the end of the inserting semi-rigid cable. Naturally, this position of the end of the semi-rigid cable has to be referenced to the proper alternative path for the necessary calibration to occur. Further, the necessary tube system includes a matrix of tubes below the reactor vessel. Normally these tubes must be removed for required below vessel service and replaced thereafter. Despite the presence of both stop valves and shear valves, the system remains as a possible escape route for water containing radioactive particles from the reactor. Further, the withdrawn cable can have mechanical complications as well as being radioactive. For these reasons, it has recently been conceived to omit the use of the TIPs, and use, instead of the TIPs, another type of reactor power measurement apparatus in combination with the LPRM system. This type of apparatus, which is referred to as a gamma thermometer, comprises a system of sensors at a fixed position in the reactor that does not require a drive mechanism, nor does it involve substantial deterioration of sensitivity. Gamma thermometers are known. In general, the gamma thermometer is a type of reactor power measurement apparatus, which detects the quantity of heat attributable to radiation, and in particular gamma rays. In contrast with a fission ionization chamber, the gamma thermometer does not, in principle, involve sensitivity deterioration. Referring to FIG. 8, a typical gamma thermometer T is illustrated in a simplistic format. Typically, the gamma thermometer T includes a metal mass 74 suspended in a cantilevered fashion within an outer tube 76. The mass of metal 74 reaches a temperature, which is directly dependent on the gamma ray flux. A reading thermocouple 78 and a reference thermocouple 80 are utilized in a series circuit. Specifically, the temperature differential between the reference thermocouple 80 (typically referenced to a temperature stable interior portion of the core) and the reading thermocouple 78 produce a voltage on paired lines 82, 84 which voltages indicate the gamma flux present which is proportional to reactor power. Gamma thermometers are placed at fixed locations within the reactor. Each gamma thermometer requires an electrical connection and readout electronics. Unfortunately, gamma ray output as measured by gamma thermometers does not provide a prompt response to power transients as required for safe operation of the reactor. In addition, gamma thermometers are expensive and the probes and associated cabling occupy a significant amount of space within the reactor. Consequently, gamma thermometers are deployed in limited numbers and provide a coarser map of the reaction rate in the core than the TIP system. Therefore, it would be desirable to overcome the above-mentioned problems associated with gamma thermometers. In one aspect of the invention, a nuclear reactor comprises a core for containing a reaction within a reactor vessel, and a plurality of instrument guide tubes extending into the core at spaced apart locations, each instrument guide tube permitting insertion and removal of a local power range monitoring string, each string having a plurality of local power range monitors for measuring a thermal neutron flux at a predetermined location within the core and an optical gamma thermometer adjacent a corresponding local power range monitor, each optical gamma thermometer comprising a metal mass having a temperature proportional to a gamma flux within the core and an optical fiber cable that measures the temperature of the metal mass, wherein the optical gamma thermometer is capable of calibrating the corresponding local power range monitor during steady state power operation of the reactor. In another aspect of the invention, an optical gamma thermometer for measuring a gamma flux in a nuclear reactor having a core, the optical gamma thermometer comprising a metal mass having a temperature proportional to a gamma flux within the core and an optical fiber cable that measures the temperature of the metal mass, wherein the optical gamma thermometer is capable of calibrating a corresponding local power range monitor during steady state power operation of the nuclear reactor. In yet another aspect of the invention, a method for calibrating a local power range monitor in a nuclear reactor, comprises: placing at least one optical gamma thermometer adjacent a corresponding local power range monitor; operating the reactor at a steady state operating condition; taking an energy balance of the reactor to determine a power output of the reactor; measuring a temperature of a metal mass of an optical gamma thermometer; correlating the power output at a portion of the reactor to the temperature measured by the optical gamma thermometer; taking a reading of the corresponding local power range monitor; and calibrating a neutron flux from the reading of the local power range monitor and the measured temperature of the optical gamma thermometer. Referring to FIG. 1, a reactor vessel V is shown with its dome removed and a local power range monitor string S in the process of insertion. Typically, the string S inserts into tube 24. The tube 24 begins at the core plate 12 and extends through the bottom of the reactor vessel V. The portion of the string S within the core extends above the tube 24. An upper portion of the string S registers to the top guide 12. Thus, the string S as ultimately held to the core stands vertically upright in measuring exposure to the neutron flux interior of the reactor core. Referring to FIG. 1, a section of a boiling water reactor vessel V is illustrated. The vessel V includes a core shroud 14 surrounding a core 15 controlled by control rods 16. In the portion of the reactor shown in FIG. 1, one or more jet pumps 13 draw water over the top of a top guide 12 downwardly in the interstitial volume between the side of the vessel V and the core shroud 14 through the jet pumps 13 to a plenum P below the core of the reactor. Water then passes upwardly through a core plate 17 into the individual fuel bundles of the core at 15. Water for the jet pumps is extracted at 19, entered at 18 to cause the required forced circulation within the reactor. Although a boiling water reactor is described herein, it will be understood that the principles of the invention can be used in any type of nuclear reactor in which in-core instrumentation is used. As shown, a single instrument guide tube 24 is illustrated for the insertion of a local power range monitor string S. It will be understood that insertion begins from above the core 15. The string S is from the top of the core 15 inserted to the bottom of the in-core guide tube 24. A seal is made at the bottom of the in-core guide tube 24. The string extends from the top of the core 15 adjacent the top guide 12. The in-core guide tube 24 can be also designed such that they can be accessed from the bottom of the reactor core 15. Each string S includes typically four monitoring sites. These monitoring sites are equally spaced between the top guide 12 and the core plate 17. They are positioned to sample four successive vertical intervals within the reactor. Although the illustration of FIG. 1 only shows one such in-core guide tube 24, it will be understood that many are in fact used to monitor a typical core 15. For example, it is not uncommon to include sixty four (64) such instrument guide tubes or about 256 discretely wired local power range monitors. It should be understood that the local power range monitors are combined in varying groups to produce required measurement. Because the combination of such groups is not pertinent to this invention and because the instrumentation for reading such local power range monitors is well known, such instrumentation will not be further described herein. Referring to FIG. 2, a local power range monitor M includes an outer cylindrical cathode 60 and an inner concentric and cylindrical anode 62 housed within an outer tube 70. Cathode 60 adjacent to anode 62 is provided with a thin coating of fissionable materials 64. Materials 64 are typically combinations of U235 and U234. As is well known in the art, U235 is expended over the life of the monitor M; the U234 breeds replacement U235 thus prolonging the in-service life of the detector M. Typically, anode 62 is mounted by insulating blocks 66 at each end so as to be concentric of the enclosing cathode 60. Preferably, an argon atmosphere 68 is present. Typically a coaxial cable leads from the detector with the center conductor connected to the anode and the outer conductor connected to the cathode. It is the direct current through the cable that provides the real time measurement of thermal neutron flux. In operation, thermal neutrons impact U235 at layer 64. Fission components dissipate into the argon gas 68 and cause electrons to flow to the anode with ions of opposite polarity to the cathode. An overall direct current is induced through the cable 69 which direct current is conventionally read. Calibration of the local power range monitor M is required because the U235 component of the layer 64 varies with in-service life. When the string S is inserted into the in-core guide tube 24 within the core 15 and the reactor is operated under steady state operating conditions, gamma thermometers can be utilized to calibrate the local power range monitors M in conjunction with a conventional reactor heat balance. Typically, the gamma thermometers are located in the vicinity of the local power range monitors M within the string S. For example, the string S may contain a total of four (4) power range detectors M at various elevations in the core 15. In one embodiment, one or more gamma thermometers may be located in the vicinity of each power range detector M to provide an indication of the gamma flux at discrete locations within the core 15. By the expedient of referencing the gamma flux to the output of the heat (energy) balance, calibration of the local power range monitors occurs over their useful in-service life. The required energy balance measuring the power output of a steady state nuclear reactor is well within the state of the art. Once this is known, measurement of the gamma thermometers can all be correlated to the intensity of the reaction at any given point. Because one or two gamma thermometers are located adjacent each and every local power range monitor M, it is thereafter possible to calibrate each local power range monitor M with the readout of its adjacent gamma thermometer. As described above, the conventional gamma thermometer T measures the temperature of the metal mass 74 by using a reading thermocouple 78 and a reference thermocouple 80. One aspect of the invention is that the temperature of the metal masses 34, 36 is not measured with the reading thermocouple 78 and a reference thermocouple 80, but with a fiber optic sensor. As used herein, a fiber optic sensor is a sensor that uses optical fiber either as the sensing element (“intrinsic sensors”), or as a means of relaying signals from a remote sensor to the electronics that process the signals (“extrinsic sensors”). Referring now to FIG. 3, a portion of the string S that includes an optical gamma thermometer, shown generally at 30, for measuring the gamma flux at discrete locations within the core 15 is shown according to an embodiment of the invention. The string S includes a cover tube having an upward flow of water therethrough. When the optical gamma thermometer is used to calibrate the local power range monitor M, it is desirable that the optical gamma thermometer 30 be located in the vicinity of the local power range monitor M. In the illustrated embodiment, the local power range monitor M is separated from the optical gamma thermometer 30 by a small distance d so that the neutron flux at the local power range monitor M is essentially uniform. For example, the distance d may be about one (1) inch. It will be appreciated that the string S may include a second optical gamma thermometer located on the opposite side of the local power range monitor M that could be separated by the same distance d, or a second distance from the local power range monitor M. As seen in FIG. 3A, the optical gamma thermometer 30 includes an optical fiber 32 that extends along the length of the string S. In the illustrated embodiment, a single gamma thermometer 30 is shown. However, several gamma thermometers can be located along the single optical fiber 32. The optical gamma thermometer 30 also includes a metal mass 34 that is thermally isolated from the ambient environment and heats in the presence of gamma flux. One way to thermally isolate the metal mass 34 from the ambient environment, for example, is to provide a gap 35 between the metal mass 34 and the outer tube 76. The optical gamma thermometer 30 also includes an optional, second metal mass 36 in thermal contact with the ambient environment that acts as a heat sink to provide a measurement of the ambient temperature to be used as a reference temperature. In an alternative embodiment, the second metal mass 36 can be omitted and the invention can be practiced without the need to measure the ambient temperature. Referring to FIG. 4, the optical fiber cable 32 includes a central fiber core 38 formed of pure or doped silica that extends along a long axis 33 and having a diameter of about 5 microns to about 100 microns. In a particular embodiment, fiber cable 32 includes a periodic modulated refractive index profile along the long axis 33, such as a cosine or Gaussian apodized refractive index profile. A fiber cladding 40 circumferentially covers the fiber core 38 and has an outer diameter of about 125 microns made from pure or doped silica. The fiber core 38 has a higher index of refraction than the fiber cladding 40 such that the fiber cable 32 acts as a waveguide for light propagation. In one embodiment, the fiber core 38 is made of pure silicon dioxide and the fiber cladding 40 is fluorine doped as a single clad fiber sensing cable. In an alternate embodiment, the fiber core 38 is F/GeO2 co-doped, and the fiber cladding 40 is F-doped as a single clad fiber sensing cable. Further, the fiber cladding 40 could be a double clad structure with a first clad lightly F-doped, and a second clad heavily F-doped. The fiber material, either pure silicon dioxide or co-doped tetrahedral O—Si—O structures, has F terminating all dangling bonds and eliminates the OH hydroxyl clusters to maintain the thermal stability and radiation resistance capability of the tetrahedral structure thermal stability. In one embodiment, a light source (not shown), such as a tunable laser, a superluminescent, light-emitting diode (LED), and the like, is positioned in light emitting communication with the optical fiber cable 32 and emits a near infrared light that propagates through the fiber core 38. It will be appreciated that the invention is not limited by the type of light source, and that the invention can be practiced using any appropriate light source. The light is transmitted or propagated through the fiber core 38 and into a photodetector or interrogator 52 (FIG. 7A). In one embodiment, the optical gamma thermometer 30 includes one or more fiber grating structures 42 that reflect light having a selected wavelength as the light is transmitted through the fiber core 38. In one embodiment, the fiber grating structure 42 comprises a fiber Bragg grating (FBG). The selected wavelength is determined by 2·n·Λ where the n is effective refractive index in the fiber core and the Λ is the grating modulation periodicity. The reflected light wavelength corresponds to a temperature of the metal masses 34, 36. The reflected light signal is transmitted to the photodetector (FIGS. 7A-7C), where the light signal is processed and/or transmitted to a computer (not shown) interfaced and/or communicating with the photodetector. For example, in one embodiment, a wireless interface 167 (FIG. 7B) may transmit electrical signals to the computer in response to light signals received by the photodetector. In one embodiment, each fiber grating structure 42 has a length along the long axis 33 of the optical fiber cable 32 of about 5 millimeters to about 20 millimeters. In the case where the fiber core 38 is made of pure silicon dioxide and the fiber cladding 40 is a double clad structure, the fiber grating structure 42 can be made from a pure quartz fiber inscribed with high-power femtosecond pulse laser grating inscription technology. In another embodiment, the fiber grating structure 42 may have a metalized cladding (not shown) surrounding the fiber cladding 40 having a polycrystalline Al, Cu/Ni, or Au coating with a thickness of about 10-20 micrometer. In the illustrated embodiment, the optical gamma thermometer 30 includes two short-period fiber Bragg gratings (FBG) 42. However, the invention can be practiced with any desirable number of FBGs. For example, the invention can be practiced with only one FBG, with three FBGs, and the like. In one embodiment, the FBG 42 has an apodized refractive index modulation along the long axis 33 with a pitch size of about 0.5 microns. It will be appreciated that the invention is not limited to a fiber Bragg grating (FBG) structure, and that the invention can be practiced with other grating structures, such as long period gratings, helical fiber, and the like. The FBG 42 is configured in a loose packaging arrangement to be effectively free from the effects of strain. In one embodiment, the length of fiber containing one or more FBGs 42 is packaged loosely in a structural cylinder, such that the outer diameter of the fiber cable 32, typically 125 microns, is slightly less than the inner diameter of the structural cylinder, typically around 140 microns. Any strain induced on the package from the outside due to thermal gradients or the mass of the structure will not be transferred to the fiber cable or FBGs. In this way, the optical response of the FBGs will be limited to temperature effects and not strain effects. The optical fiber cable and the FBGs 42 may also be coated with a thin layer of material, such as aluminum, to protect the fiber cable from damage within the structural cylinder. The strain effect on the FBGs 42 due to differences in thermal expansion coefficients between the glass fiber and the coating can be measured or calculated, and factored into the translation between optical wavelength shift and temperature. Light is reflected at a single wavelength from the fiber Bragg grating structures 42. The reflected signal is a function of material properties and grating structure, such as the index of refraction (n), the grating modulation number N, the normalized mode number V, the grating period Λ, and the grating length LG. The thermal induced wavelength shift, reflected power loss, and Bragg peak resonant width from the FBG 42 can be described as: Δ λ ( T ) = λ B ( 1 n ∂ n ∂ T + 1 Λ ∂ Λ ∂ T ) Δ T ; R = tanh 2 [ π Δ nL G λ B ( 1 - 1 V 2 ) ] ; Δ λ = λ B ( Δ n 2 n ) 2 + ( 1 N ) 2 Obviously, the relative wavelength shift is proportional to gamma ray induced temperature change. The parameters 1 n ∂ n ∂ T and 1 Λ ∂ Λ ∂ T are thermo-optic coefficient and coefficient of thermal expansion, determined by the fiber material properties. Referring now to FIG. 5, an example of a reflection response to an optical gamma thermometer with an array of ten (10) fiber Bragg gratings is shown. As shown in FIG. 5, the reflected power is in the range between about −15 dBm to about −50 dBm for wavelengths between about 1510 nm to about 1590 nm. The separation of the peaks varies as a function of the grating period, and the modulation of the center wavelength of each peak is an indication of the gamma flux (or temperature). Temperature resolution is less than 1 degree C. Optical fiber, by the nature of its material properties, is an ideal distributed temperature sensor. There are several methods available for extracting distributed temperature information from optical fiber. These include techniques based on Raman, Brillouin, and Rayleigh scattering, as well as those involving multiplexed fiber Bragg gratings (FBGs). Techniques based on Raman and Brillouin scattering use time-of-flight (TOF) of optical pulses to resolve position. Due to the limits of how short optical pulses can be made and detected with good noise performance, the systems are thus limited in spatial resolution from about 0.1 m to about 1 m. Methods that employ FBGs can achieve higher resolution, but are often limited by the number of gratings in each fiber. Whenever more than 50 FBGs in each fiber are needed, a multiplexing method, either with an optical coupler/combiner or optical switch, could be used for multiple FGB sensor interrogation. Both FBG-based and scatter-based techniques may require a special optical fiber, for example, a radiation resistant optical fiber, and the like. Raman scattering allows one to measure temperature. The signal that one obtains is typically a spectrum whose peaks are linearly related to the material symmetry and structural properties. These peaks occur at characteristic intervals that depend on the physical characteristics of the optical phonon vibration, thus producing a fingerprint unique to that material and making this a good method for material analysis. In Raman spectroscopy, that characteristic interval is the frequency shift from the optical phonon vibration modes. This frequency shift is related to the rotational and vibrational components of each phonon excitation energy at the time it encounters the laser, and it appears as a positive shift (Stokes scattering) when the phonons receive energy from the laser and a negative shift (anti-Stokes scattering) when the phonons emit energy. The relative intensity of the Stokes and anti-Stokes peaks depends on the temperature that a system of optical phonon finds itself in, which follows a Boltzmann distribution. Brillouin scattering occurs when light in a medium (such as water or a crystal) interacts with time dependent density variations and changes its energy (frequency) and path. The density variations may be due to transient dielectric function fluctuations. Whenever the light interaction with transverse and longitude acoustic waves, the scattered light signal also contains Stokes and Anti-Stokes signal components. However, the determination of the temperature is by frequency shift of a specific acoustic wave, or velocity variation from transverse or longitude wave. For isotropic fiber material, the velocity of the transverse and longitude is determined by (C44/ρ)1/2 and (C11/ρ)1/2 where C11 and C44 are the elastic module and ρ is fiber material mass density. The determination of the temperature is by the measured velocity or frequency shift. The inelastic scattering process of Brillouin light scattering is in principle the same as Raman scattering. Historically, Brillouin scattering denominates the scattering of acoustic phonons, while Raman scattering refers to the scattering from molecule vibrations and optic phonons. Nowadays, the difference between Brillouin scattering and Raman scattering is considered to lie in the different experimental techniques and the resulting different available frequency range. The term Brillouin scattering labels an experimental detection of the frequency shift with an interferometer, while Raman scattering labels a setup employing a grating spectrometer. Brillouin scattering is technically limited to the detection of quasiparticles with frequencies below about 500 GHz, while with Raman scattering much higher frequencies in the THz range can be measured. For intense beams travelling in a medium, such as an optical fiber, the variations in the electric field of the beam itself may produce acoustic vibrations in the medium via electrostriction. The beam may undergo Brillouin scattering from these vibrations, usually in opposite direction to the incoming beam, a phenomenon known as stimulated Brillouin scattering (SBS). Typical frequency shifts of longitudinal and transverse acoustic modes are of the order of about 20-50 GHz. Stimulated Brillouin scattering is one effect by which optical phase conjugation can take place. For all above sensing methods, either based on FBGs or on fiber itself, the fiber sensing cable requires specialty fiber material that can strongly resist high-energy particles or quasi-particles damage such as gamma-ray, neutrons, and alpha and beta rays etc. The required single mode or multimode fibers are composed of fiber core, either pure silicon dioxide or doped silicon dioxide, fiber cladding, either pure silicon dioxide when the fiber core is co-doped with dopant such as F/GeO2 or doped with fluorine when the fiber core is pure silicon dioxide. In one embodiment, such a fiber is composed of pure silicon dioxide core of 4-62.5 μm in diameter, and fluorine doped fiber cladding of 125 μm in diameter. In the other embodiment, the fiber cladding is made of two cladding structures. First cladding is about 24-30 μm for single mode fiber and 82.5 μm for multimode fiber with light fluorine doping concentration; second cladding area, from above cladding diameter to 125 μm, has a higher fluorine doping concentration so that the refractive index profile of the fiber is reduced at from 10−4 to 10−2 for reduced transmission loss. The fiber material, silicon dioxide, is preferred to have condensed, tetrahedral O—Si—O structures such that the dangling bonds are terminated mainly by fluorine. The impurity levels in the fiber material band-gap have to be eliminated such as by a thermal annealing process that effectively enlarges the doped fiber material band-gap. Preconditioning suppresses the carriers from transporting between the covalence and conduction bands, thereby minimizing any changes in conductivity or refractive index. On the other hand, the substitution of the O—Si—OH bond by O—Si—F could eliminate the OH hydroxyl clusters, which can cause fiber darkening. Because a metalized coating layer, such as Al, Cu/Ni, Au, is used for hermetical sealing and package, its loose microstructure has to be modified to have polycrystalline morphology through thermal annealing at elevated temperature for certain time. The desired temperature for the fiber pre-conditioning process is less than 400° C. for Al coated fiber and fiber sensors, and less than 500° C. for Cu/Ni, and less than 700° C. for Au coating. The desired time should be longer than 24 hours in an inert environment. Referring now to FIG. 7A, a system, shown generally at 150, for determining a gamma flux with multiple optical gamma thermometers 30 at multiple, discrete locations in the reactor core 15 from a single optical fiber cable 32, is shown. For example, the string S may have eight (8) optical gamma thermometers 30 at discrete locations along the optical fiber cable 32. The system 150 includes an interrogator, shown generally at 152, in optical communication with an optical fiber cable 32. In this embodiment, the optical gamma thermometers 30 include the fiber Bragg grating structure 42 to measure the temperature of the metal mass 34 and/or 36 at discrete locations in the reactor core 15. The interrogator 152 includes a light source 154 and a light receiver 156, and a processor and communications unit 158. An optical coupler or circulator 159 receives the light transmitted from the light source 154 and transmits a portion of the light through fiber core 38 of the optical fiber cable 32. The light detector 156 receives the light reflected from the optical gamma thermometers 30 through the optical coupler or circulator 159. The processor 158 receives a signal of the reflected light from the light detector and processes the signal in accordance with techniques known in the art. Referring now to FIG. 7B, a system, shown generally at 160, for determining a gamma flux with multiple optical gamma thermometers 30 at multiple locations in the reactor core 15 from multiple optical fiber cables 32, is shown. For example, the system 160 may include sixty-four (64) strings S having eight (8) optical gamma thermometers 30 in each string S. Similar to the embodiment shown in FIG. 7A, the optical gamma thermometers 30 include the fiber Bragg grating structure 42 to measure the temperature of the metal mass 34 and/or 36 at discrete locations in the reactor core 15. The system 160 includes an interrogator, such as a FBG interrogator, shown generally at 162, in optical communication with an optical splitter or switch 164. Similar to the interrogator 152 shown in FIG. 7A, the interrogator 162 includes a light source 154 and a light receiver 156, and a processor and communications unit 158. The interrogator 162 is in communication with a Network interface 166 for transmitting data to a processor 168, such as a laptop computer, and the like. The Network interface 166 can also transmit to a wireless interface 167 for wireless transmission to the processor 168. Referring now to FIG. 7C, a system, shown generally at 170, for determining a gamma flux with multiple optical gamma thermometers 30 at multiple locations in the reactor core 15 from multiple optical fiber cables 32, is shown. Unlike the embodiments shown in FIGS. 7A, 7B, the optical gamma thermometers 30 do not include the fiber Bragg grating structure 42 to measure the temperature of the metal mass 34 and/or 36, but measure the temperature continuously along the length of the optical fiber cable 32 in the reactor core 15. A continuous sensor interrogator 172 is in optical communication with an optical switch 174. Similar to the interrogator 52 shown in FIG. 7A, the interrogator 172 includes a light source 154 and a light receiver 156, and a processor and communications unit 158. Similar to the interrogator 162 shown in FIG. 7B, the interrogator 172 is in communication with a Network interface 176, which in turn, may be in communication with a wireless transmitter 167 for transmitting data to a processor 168, such as a laptop computer, and the like. It will be appreciated that the optical gamma thermometer of the invention does not include portions thereof, which, with increased in-service life, have decreasing effectiveness. This being the case, it will be understood that with reference to any heat balance, the expectancy is that the output of the optical gamma thermometers 30 will remain substantially unchanged. As described above, the optical gamma thermometer of the invention can provide a measurement of the temperature as a result of gamma flux in an in-core instrument tube using a single optical fiber cable, rather than with the several cables that are required with the conventional gamma thermometer that uses thermocouples to measure temperature. As a result, the optical gamma thermometer of the invention occupies much less space within the in-core instrument tube and costs much less than the conventional gamma thermometer. In addition, a larger number of temperature measurements in a single optical fiber cable are possible, providing a much more accurate three-dimensional temperature map (gamma flux) of the reactor core. Further, the optical gamma thermometer can be used to calibrate computer simulations relating to the depletion of nuclear fuel during a fuel cycle to better optimize refueling plans. Even further, loose tube packaging of the optical gamma thermometer of the invention with a fiber Bragg grating (FBG) helps eliminate false readings due to the effects of strain. Still even further, a system using the optical gamma thermometer of the invention can replace the conventional TIP system to significantly reduce cost. While the invention has been described with reference to an exemplary embodiment, it will be understood by those skilled in the art that various changes may be made and equivalents may be substituted for elements thereof without departing from the scope of the invention. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the invention without departing from the essential scope thereof. Therefore, it is intended that the invention not be limited to the particular embodiment disclosed as the best mode contemplated for carrying out this invention, but that the invention will include all embodiments falling within the scope of the appended claims. |
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046738142 | summary | FIELD OF THE INVENTION The invention relates to a container for receiving and safely storing radioactive materials and other materials damaging to living organisms. The container is especially suited for storing vitrified radioactive fission products or irradiated nuclear reactor fuel elements. The container includes a vessel and a sealing cover. One end of the vessel is provided with a circular opening into which the sealing cover can be placed to tightly seal the container. BACKGROUND OF THE INVENTION Radioactive materials such as vitrified radioactive fission products or irradiated nuclear fuel elements have to be placed in special containers for the purpose of transport and storage. The containers have a high shielding to radiation and a sufficient cooling surface as well as a high stability. Metal containers guarantee a safe enclosure for radioactive waste products. The metal containers are open only at one end and it is this opening through which the materials to be stored are passed. After filling a container destined for terminal storage with glass from highly radioactive fission products, it has been conventional practice to place a cylindrical cover in the open end of the vessel and to weld the cover to the abutting container rim. By means of the tight seal of the vessel with the sealing cover, it was believed that the radioactive materials or materials damaging to living organisms could be safely separated from the ambient. The results obtained up until now with this procedure have been unsatisfactory. The welding activity has to be carried out in a so-called "hot cell." Accordingly, it was necessary to conduct the welding operation from a remote location with the aid of remotely-controlled apparatus. Up to now, the desired impermeability of the seal to gas of 10.sup.-3 Torr Liter/Second could not be obtained with certainty and reproducibility. The operating person carrying out the welding operation must necessarily perform each welding operation individually and always with a different quality. SUMMARY OF THE INVENTION It is an object of the invention to provide a container of the kind described above wherein a safe closure of the container is achieved by the sealing cover with the required impermeability to gas. It is a further object of the invention to provide method for tightly sealing the container. The container includes a vessel and a sealing cover for closing the vessel off from the ambient. According to a feature of the invention, the open end of the inner bore of the vessel is widened to define a conical surface whereat the sealing cover is seated when pressed into the vessel, the sealing cover having an outer peripheral conical surface formed to converge toward the interior of the vessel and the outer peripheral conical surface having a taper corresponding to the taper of the conical surface of the vessel. The wall of the vessel and the sealing cover are welded together by means of a fused-mass extending about the periphery of the sealing cover. After the vessel is filled, the conical sealing cover is pressed into the conical opening of the vessel. As the conical sealing cover is pressed into the vessel, the sealing cover exerts a radial force on the vessel wall surface so that irregularities of this surface surrounding the cover are minimized and even eliminated. The surface of the vessel wall adapts itself well to the outer peripheral surface of the sealing cover. By means of the conical configuration of the two surfaces which are pressed against each other, namely, the conical surface of the vessel and the peripheral surface of the sealing cover, a considerable improvement in the seal between the inner wall of the vessel and the sealing cover is obtained. The conical sealing surfaces are responsible for the improvement in the seal of the connection between sealing cover and vessel. After completing the fused-mass welding operation, a safe and tight closure of the container is obtained. In an advantageous further embodiment of the invention, the upper outer edge of the sealing cover lies beneath the rim of the vessel whereby the vessel rim and the upper edge of the sealing cover are welded together with a fillet weld extending about the periphery of the sealing cover. If the conical sealing cover is pressed so deep into the conical section of the container so that there is sufficient room between the upper edge of the sealing cover and the rim of the vessel to accommodate a fillet weld, then the required impermeability to gas and mechanical joint between the vessel and sealing cover can be obtained. According to another feature of the invention, the upper portion of the outer peripheral surface of the cover can be bevelled so that the same diverges away from the remainder of this peripheral surface. With this short cylindrical bevel of the cover, a welding starter gap of wedge-shaped configuration is obtained between the inner wall of the vessel and the peripheral surface of the sealing cover. After pressing the sealing cover into the conical seat of the vessel, the sealing cover is welded to the vessel by means of a fused-mass weld under the application of shielding gas. In this way, the shielding gas can blow the melt into the annular wedge-shaped gap. This embodiment is especially suited for a remotely-controlled welding process which can be conducted automatically in a hot cell. The quality of the weld joint is higher than if an operating person individually conducted each welding operation. Further, the weld contemplated by this embodiment of the invention is reproducible every time. In still a further embodiment of the invention, the sealing cover is provided with a valve accessible from the outside and which valve communicates with the interior of the vessel; this arrangement permits the valve to be connected to a test gas source. By holding the valve open, an equalization of pressure is possible during the welding process between the space defined by the sealing cover and vessel and the space surrounding the system. On the other hand, after the welding process has been terminated, a simple test of the impermeability to gas of the weld joint is possible. A further advantageous embodiment of the invention is achieved by mounting a valve in a recess of a projection formed on the sealing cover. The recess defined by the projection constitutes a protective chamber for the valve so that the latter will not become damaged should the container be inadvertently dropped. Another advantageous embodiment of the invention is to provide a plug for closing off the recess. The plug can, for example, be configured to threadably engage an internal thread of the recess. Where the parts are made of metal and the container receives radioactive waste materials, it is preferable to weld the plug with the projection. According to a further embodiment of the invention, the projection can be configured as a cylinder and adapted to threadably engage the sealing cover in a centrally disposed threaded bore formed in the latter. The projection can furthermore be given the shape of a knob. This embodiment permits the operation of sealing the container to be conducted without difficulty by remotely-controlled programmed machines. The invention also is directed to a method for tightly sealing a container for receiving and safely storing radioactive materials and other materials dangerous to living organisms especially such materials as vitrified radioactive fission products or irradiated nuclear reactor fuel elements, and wherein the container includes a vessel having a circular opening at one end thereof for receiving the materials to be stored and which opening is closed off from the ambient with a sealing cover placed therein. The method of the invention includes the steps of: conically widening the inner bore of the vessel at the opening end thereof to define a conical seating surface; turning the outer peripheral surface of the cover to have a conical surface having the same taper as the taper of the seating surface; turning the upper portion of the outer peripheral surface of the cover to define a cylindrical surface; pressing the cover down onto the conical seating surface to a depth below the rim of the vessel after filling the vessel with the materials to be stored thereby defining an annular groove of wedge-shaped section; joining the cover to the vessel by means of a gas-shielded arc weld while maintaining an equalization of pressure between the interior of the container and the ambient, the flow of shielding gas being directed from above into the annular gap of wedge-shaped section and, the weld formed in this manner including: a first portion defining an annular bevel weld filling in the annular groove of wedge-shaped section; and, a second portion defining an annular fillet weld disposed in the fillet defined by the top peripheral edge surface of the cover and the remainder of the conical seating surface of the vessel above the bevel weld; and discontinuing the maintenance of the equalization of pressure after completing the welding step. With the aid of the invention, containers for receiving material which is radioactive or dangerous to living organisms can, after they have been filled, be safely sealed with a high impermeability to gas and again be tested as to the integrity of the seal. The invention permits the utilization of remotely-controlled programmed robots and automatic welding equipment to produce connections of a reproducible high quality. |
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042319765 | claims | 1. A process for the production of a plutonium-uranium ceramic fuel wherein green pellets are formed by pressing from: (i) a mixture of (a) an oxide, finely-divided carbon-containing oxide, carbide, nitride, oxycarbide, or carbonitride of plutonium, in the form of microspheres obtained by a wet chemical process, and (b) a like compound of uranium in a form selected from such microspheres and fine-grained to pulverulent materials; or (ii) a like compound of a plutonium-uranium solid solution of which at least 1% is plutonium, in the form of microspheres obtained by a wet chemical process, alone or in admixture with a material selected from the said plutonium-compound microspheres, the said uranium-compound microspheres, and the said fine-grained to pulverulent materials, (a) mixing together a plutonium-containing compound and a uranium-containing compound, said plutonium-containing compound being in the form of microspheres produced by a wet chemical process and being selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride; and said uranium-containing compound being in the form of either microspheres or else a fine-grained to pulverulent form of a uranium compound selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, (b) pressing the mixture formed in step (a) to form green pellets, and (c) heat treating the green pellets in step (b) to form a plutonium-uranium ceramic fuel. (a) forming microspheres of a plutonium-uranium solid solution by a wet chemical process, said plutonium-uranium solid solution containing at least 1% plutonium, said plutonium-uranium solid solution containing plutonium compounds selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, and said plutonium-uranium solid solution containing uranium compounds selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, (b) pressing the microspheres in step (a) to form green pellets, and (c) heat treating the green pellets in step (b) to form a plutonium-uranium ceramic fuel. (a) mixing together (i) microspheres obtained from a wet chemical process of a plutonium-uranium solid solution, said solid solution containing at least 1% plutonium, said plutonium-uranium solid solution containing plutonium compounds selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, and said plutonium-uranium solid solution containing uranium compounds selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, (ii) microspheres of plutonium-containing compounds produced by a wet chemical process, said plutonium-containing compounds selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, and (iii) a uranium-containing compound either in microsphere form or else a fine-grained to pulverulent form, said uranium-containing compound selected from the group consisting of an oxide, a carbon-containing oxide, a carbide, a nitride, an oxycarbide, and a carbonitride, (b) pressing the mixture formed in step (a) to form green pellets, and (c) heat treating the green pellets in step (b) to form a plutonium-uranium ceramic fuel. 2. A process according to claim 1 wherein, for the production of sintered pellets with a specified plutonium to heavy metal ratio, xerogel microspheres of a solid solution of uranium plutonium obtained in a gelation process are pressed to green pellets with the same plutonium to heavy metal ratio and the green pellets are calcined at a certain temperature and sintered at a higher temperature in the course of a heat treatment. 3. A process according to claim 2, wherein xerogel microspheres produced by a gel-precipitation process are used for the production of uranium-plutonium mixed oxide pellets. 4. A process according to claim 3, wherein the xerogel microspheres of plutonium and uranium have mixed or uniform diameter of 20 to 100 .mu.m. 5. A process according to claim 2, wherein the green pellets are calcined in a reducing gas of 80% argon and 20% hydrogen at 500.degree. C. and subsequently are sintered in an oven at 1400.degree. to 1600.degree. C. 6. A process according to claim 5, wherein the green pellets are sintered to 95-98% of the theoretical density. 7. A process according to claim 2, wherein xerogel microspheres of uranium-plutonium made according to the "EIR sol-gel process" are used for the production of uranium-plutonium mixed carbide pellets, wherein the microspheres contain about 12 weight parts, related to the total weight, of finely divided carbon-black and have diameters lying in the range of 20 to 200 .mu.m. 8. A process according to claim 7, wherein the green pellets are calcined in flowing gas of 20% hydrogen and 80% argon at about 500.degree. C. and then are subjected to a heat treatment in flowing argon of 1700.degree. to 1800.degree. C., during which a carbon reduction takes place and the pellets are sintered. 9. A process according to claim 8, wherein the pellets are sintered to 95% of the theoretical density. 10. A process according to claim 1, wherein for the production of sintered pellets with a specified plutonium to heavy metal ratio, a xerogel microsphere of uranium obtained in a gelation process or fine-grained to pulverulent uranium material is used in a weight ratio determined by the plutonium to heavy metal ratio and are mixed with xerogel microspheres of plutonium obtained in a gelation process, the uranium particles and the plutonium microspheres having at least approximately the same size distribution and about the same mean diameter, and wherein the particle mixture is pressed to green pellets, and in the course of a heat treatment the green pellets are calcined at a specified temperature and sintered at a higher temperature. 11. A process according to claim 10, wherein the uranium microspheres or the fine-grained to pulverulent material and the plutonium microspheres pressed to green pellets have particle diameters lying in the range of 5 to 25 .mu.m and a mean diameter of 10 to 15 .mu.m. 12. A process according to claim 10, wherein uranium xerogel microspheres made by the "Julich H-Process" and plutonium xerogel microspheres produced by an "EIR Sol-Gel Process" are used for the production of uranium-plutonium mixed oxide pellets, particularly for LW reactors. 13. A process according to claim 12, wherein the green pellets are calcined in a reducing gas of 80% argon and 20% hydrogen at about 500.degree. C. and then sintered in argon at 1400.degree. to 1600.degree. C. 14. A process according to claim 13, wherein the pellets are sintered to 95-98% of the theoretical density. 15. A process according to claim 1, wherein the calcined xerogel microspheres of plutonium obtained in a gelation process and the calcined xerogel microspheres of uranium obtained in a gelation process or a corresponding fine-grained to pulverulent uranium material are mixed together in a certain weight ratio, the uranium particles and the plutonium microspheres having at least approximately the same size distribution and about the same mean diameter, and wherein the mixture is pressed to green pellets and the green pellets are sintered at a higher temperature. 16. A process according to claim 15, wherein the uranium particles and the plutonium microspheres have a diameter in the range of 5 to 25 .mu.m. 17. A process according to claim 15, wherein uranium oxide microspheres produced according to an "ORNL Sol-Gel Process" and plutonium oxide microspheres produced by a PuO.sub.2 Sol-gel process are used for the production of uranium-plutonium mixed oxide pellets, particularly for FB reactors. 18. A process according to claim 15, wherein the green pellets are sintered in a flowing gas of argon and 4% hydrogen at 1200.degree. to 1300.degree. C. 19. A process according to claim 18, wherein the pellets are sintered to 95-98% of the theoretical density. 20. A process according to claim 1, wherein the plutonium xerogel microspheres obtained in a gelation process are calcined at a specified temperature and the thus obtained plutonium oxide microspheres are subjected to reaction sintering at a higher temperature; that the sintered plutonium microspheres are mixed with microspheres or uranium with a fine-grained to pulverulent uranium material, both kinds of particles having at least approximately the same size distribution and about the same mean diameter, and wherein green pellets are pressed from the mixture and the green pellets are subjected to final sintering. 21. A process according to claim 20, wherein the uranium xerogel microspheres obtained in a gelation process are calcined at a specified temperature and the thus obtained uranium microspheres are subjected to reaction sintering at a higher temperature and the sintered uranium oxide microspheres are mixed with the reaction-sintered plutonium microspheres. 22. A process according to claim 20, wherein the reaction-sintered plutonium microspheres are mixed with a fine-grained to pulverulent uranium material which is calcined and/or reaction sintered. 23. A low-dust process for the production of a homogeneous plutonium-uranium ceramic fuel which comprises the steps of 24. The low-dust process in accordance with claim 23 wherein said uranium-containing compound is in the form of microspheres, and wherein both said plutonium-containing compound microspheres and said uranium-containing compound microspheres are separately calcined prior to being mixed together in said step (a). 25. The low-dust process in accordance with claim 24 wherein said uranium-containing compound microspheres include finely divided carbon. 26. The low-dust process in accordance with claim 25 wherein prior to the mixing together of said plutonium-containing compound microspheres and subsequent to the separate calcinations of said microspheres, said microspheres are separately presintered. 27. The low-dust process in accordance with claim 23 wherein said uranium-containing compound is in the form of a fine-grained to pulverulent compound, and wherein said plutonium-containing compound is calcined prior to being mixed together with said uranium-containing compound in step (a). 28. The low-dust process in accordance with claim 23 wherein said uranium-containing compound is in the form of a fine-grained to pulverulent compound, and wherein said plutonium-containing compound microspheres are pre-sintered prior to being mixed together with said uranium-containing compound in said step (a). 29. The low dust process in accordance with claim 28 wherein said uranium-containing compound is in the form of a fine-grained to pulverulent compound is pre-sintered prior to being mixed with said pre-sintered plutonium-containing compound. 30. A low-dust process for the production of a homogeneous plutonium-uranium ceramic fuel which comprises the steps of 31. A low-dust process for the production of a homogeneous plutonium-uranium ceramic fuel which comprises the steps of |
claims | 1. A method for redundant state determination of positively charged particles of a particle beam, comprising the steps of:determining a first state of the positively charged particles using an extraction process of the positively charged particles from a synchrotron;measuring a second state of the positively charged particles using a signal generated in a beam transport path downstream from said synchrotron;communicating both the first state and the second state to a treatment delivery control system. 2. The method of claim 1, said step of measuring the second state further comprising the step of:using a scintillation detector of a tomography system to determine the second state. 3. The method of claim 2, said step of determining the first state further comprising the step of:measuring a secondary electron emission from an extraction material upon transmission of the positively charged particles therethrough, wherein both the first state and the second state represent a number of the positively charged particles per unit time. 4. The method of claim 3, further comprising the step of:calculating a third state of the positively charged particles using an applied field state. 5. The method of claim 4, further comprising the step of:determining energy of the positively charged particles using a period of the applied field state to calculate an energy of the positively charged particles while circulating in said synchrotron. 6. The method of claim 4, said applied field state comprising:an applied magnetic field through a magnet, the positively charged particles passing through the magnet. 7. The method of claim 6, further comprising the step of:using photons emitted from a sheet upon transmission of the positively charged particles through the sheet to determine a fourth state of the positively charged particles. 8. The method of claim 7, said step of using photons resulting in a first determined position of the positively charged particles and said step of using a scintillation detector resulting in a second determined position of the positively charged particles. 9. The method of claim 7, said first state, said second state, said third state, and said fourth state determined using an extraction process, scintillation process, calculation process, and photonic process, respectively. 10. The method of claim 6, said step of measuring a second state using said scintillation detector resulting in a first determined energy of the positively charged particles and said step of calculating a third state resulting in a second determined energy of the positively charged particles. 11. The method of 10, said step of determining a first state of the positively charged particles using an extraction process resultant in a third determined energy of the positively charged particles. 12. The method of claim 2, said step of using the scintillation detector further comprising:determining an intensity of the positively charged particles using a magnitude of emitted photons upon the positively charged particles striking a scintillation material of the tomography system. 13. The method of claim 12, said step of using the scintillation detector further comprising:determining a position of the positively charged particles using a position of emitted photons upon the positively charged particles striking a scintillation material of the tomography system. 14. The method of claim 12, the intensity of the positively charged particles comprising a number of the positively charged particles per unit time. 15. The method of claim 2, said step of determining the first state of the positively charged particles further comprising the step of using a signal generated during extraction of the positively charged particles, said step of measuring the second state of the positively charged particles further comprising the step of using photons emitted from a sheet upon transmission of the charged particles therethrough. 16. The method of claim 2, said step of measuring the second state using a signal generated in a beam transport path after passage through the patient. 17. The method of claim 1, said step of determining a first state generating a first intensity of the positively charged particles further comprising the step of using electrons emitted from a material upon transmission of the charged particle beam through said material, said step of measuring the second state further comprising the step of generating a second intensity of the positively charged particles using photons emitted from a sheet upon transmission of the charged particle beam through said sheet. 18. An apparatus for redundant state determination of positively charged particles of a particle beam, comprising:a treatment delivery control system communicatively linked to a synchrotron, said treatment delivery control system configured to determine a first state of the positively charged particles using an extraction process of the positively charged particles from said synchrotron;said treatment delivery control system configured to receive a signal generated in a beam transport path downstream from said synchrotron, said signal measuring a second state of the positively charged particles,wherein said treatment delivery control system receives communication of both the first state and the second state. 19. The apparatus of claim 18, said treatment delivery control system further comprising:a current to voltage converter, a current to the converter comprising a secondary emission of electrons resultant from the positively charged particles passing through an extraction material, a voltage from said converter used to measure the first state of the positively charged particles; anda scintillation detector configured to detect the signal resultant from the positively charged particles striking a scintillation plate of a tomography system. 20. The apparatus of claim 19, said treatment delivery control system further comprising:a photon sensor configured to detect photons emitted from a sheet upon transmission of the positively charged particles through the sheet, the photon sensor used to determine a third state of the positively charged particles comprising at least one of a position and an intensity of the positively charged particles. |
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description | This application is a US 371 application from PCT/RU2018/000565 filed Aug. 28, 2018, which claims priority to Russian Application No. RU 2018117551 filed May 11, 2018, the technical disclosures of which are hereby incorporated herein by reference. The invention relates to devices for eliminating radioactive contamination of radioactive waste, namely, plants for electrochemical decontamination of metal radioactive waste. The plant for electrochemical decontamination of the upper duct carrier of uranium-graphite nuclear reactors is known (the RF patent for invention No. 2096845) that comprises a water supply pipe, a cathode placed in an electrolyte-filled bath and connected to a DC source, an anode also connected to a DC source and connected to a treated metal radioactive waste item. The disadvantages of the known plant, besides limiting the possibility of decontaminating only structural elements of channel-type reactors of the RBMK 1000 and 1500 type, are also the absence of the process automation and the need for a certain preparation of the decontamination solution, the need for additional quality control of the decontamination solution during the electrochemical decontamination process, the need for further collection of the spent decontamination solution, its recycling or disposal, as well as the need to provide exhaust ventilation of the bath or of the device itself. The closest technical solution chosen as a prototype is a device for electrochemical decontamination of metal surfaces (the RF patent for utility model No. 127237) that comprises a replaceable electrode, bundles of electrically conductive material fastened in it that are in contact with the surface being decontaminated, electrolyte distribution chamber, waste electrolyte collection chamber and gas collection chamber, whilst each of the chambers is equipped with at least one nozzle, wherein the electrode is connected with the electrolyte distribution chamber. Despite the fact that this plant allows achieving a higher current density on the electrodes at equal voltage, its disadvantages are the need to provide the plant with a prepared decontamination solution, the need to control the quality of the decontamination solution during the electrochemical decontamination process and the need for further processing or disposal of the spent decontamination solution. The task of the claimed invention is to expand the functionality of the plant and increase the efficiency of the use of decontamination solution. The technical result achieved by the claimed invention is to provide adaptive processing of the decontamination solution for reuse with a simultaneous increase of the decontamination solution processing speed and improvement of its quality for reuse. The said technical result is achieved due to the fact that the plant for electrochemical decontamination of metal radioactive waste includes a pipe equipped with shut-off valves, a radioactive waste processing module that comprises a unit for electrochemical decontamination of metal radioactive waste connected by a ventilation channel to the ventilation module and pipe for decontamination solution supply and discharge equipped with shut-off valves and, with the module for decontamination solution receiving. According to the claimed solution, the plant is equipped with a decontamination solution preparation module connected with a pipe for decontamination solution supply and discharge, equipped with at least one pump, with a unit for electrochemical decontamination of metal radioactive waste and a module for decontamination solution receiving, whilst the module for decontamination solution receiving is equipped with devices for cleaning and pH correction of decontamination solution, and the unit for electrochemical decontamination of metal radioactive waste, the module for decontamination solution receiving and the decontamination solution preparation module are equipped with pH measurement elements. Wherein the unit for electrochemical decontamination of metal radioactive waste may include a cylindrical working container with a lower conical part, a basket for the metal radioactive waste treated element inside the working container, a high-pressure water supply pipe and a pipe for decontamination solution supply and discharge equipped with shut-off valves connected with the working container, a DC source, the negative output of which is connected according to the cathode circuit, and a positive output is connected according to the anode circuit with the basket for the metal radioactive waste treated element, a mixing device connected with the working container, a precipitation discharge unit located in the lower part of the working container, and an ionizing radiation power monitoring device, a temperature monitoring device and a decontamination solution level monitoring device that are connected with the working container. In addition, the module for decontamination solution receiving may include a decontamination solution processing container connected with the pipe for decontamination solution supply and discharge equipped with an ion-selective filter, and an alkaline substance supply pipe equipped with shut-off valves, the ionizing radiation power monitoring device, the mixing device, the precipitation discharge device located in the lower part of the decontamination solution processing container, and the decontamination solution level monitoring device located inside the decontamination solution processing container, that are connected with the decontamination solution processing container. The decontamination solution preparation module may include a decontamination solution preparation container that is connected with supply pipes for water and acidic substances, and the pipe for decontamination solution supply and discharge that are equipped with shut-off valves, a mixing device connected with decontamination solution preparation container, and the decontamination solution level monitoring device installed inside the decontamination solution preparation container. Primarily, the ventilation module includes a ventilation duct, a water separator and a hydrogen post-combustion device that are installed in it. The radioactive waste processing module may include a metal radioactive waste degreasing unit equipped with a container for metal radioactive waste degreasing that is connected with a degreasing solutions supply pipe, high-pressure water, steam and air supply pipes equipped with shut-off valves, a pH measuring device and a precipitation discharge device connected with the container for metal radioactive waste degreasing. In addition, the radioactive waste processing module may include an etching unit for non-metal coatings of metal radioactive waste equipped with a container for etching of non-metal coatings of metal radioactive waste that is connected with acidic substances supply pipe, high-pressure water and air supply pipes equipped with shut-off valves, a pH measuring device and a precipitation discharge device connected with the container for etching of non-metal coatings of metal radioactive waste. Also, the radioactive waste processing module may include a flushing unit connected by a high-pressure water supply pipe equipped with shut-off valves with the unit for electrochemical decontamination of metal radioactive waste, the metal radioactive waste degreasing unit, and the etching unit for non-metal coatings of metal radioactive waste. In addition, the decontamination solution preparation module, the module for decontamination solution receiving, the metal radioactive waste degreasing unit and the etching unit for non-metal coatings of metal radioactive waste can be connected with the ventilation module by a ventilation duct. The pipe for decontamination solution supply and discharge can be designed to transfer the decontamination solution from the decontamination solution preparation module to the unit for electrochemical decontamination of metal radioactive waste and the module for decontamination solution receiving, from the unit for electrochemical decontamination of metal radioactive waste to the module for decontamination solution receiving, from the module for decontamination solution receiving to the unit for electrochemical decontamination of metal radioactive waste. The claimed invention is shown in the drawing that explains the invention. The FIGURE shows the pneumatic-hydraulic diagram of the plant for electrochemical decontamination of metal radioactive waste. The claimed technical solution is a plant for electrochemical decontamination of metal radioactive waste, is explained by a specific design described below, however, the given example is not the only possible, but it clearly demonstrates the possibility of achieving this set of essential features of the claimed technical result. The plant for electrochemical decontamination of metal radioactive waste includes a radioactive waste processing module 1 that comprises a unit for electrochemical decontamination of metal radioactive waste connected by a ventilation duct 2 with a ventilation module 3. In addition, the plant for electrochemical decontamination of metal radioactive waste includes a pipe equipped with shut-off valves, which are valves, taps, gate valves, damper valves. In addition, the unit for electrochemical decontamination of metal radioactive waste is connected by a pipe 4 for decontamination solution supply and discharge equipped with shut-off valves, with a module 5 for decontamination solution receiving. The plant for electrochemical decontamination of metal radioactive waste is also equipped with a decontamination solution preparation module 6 that is connected by the pipe 4 for decontamination solution supply and discharge with the unit for electrochemical decontamination of metal radioactive waste and the module 5 for decontamination solution receiving. The pipe 4 for decontamination solution supply and discharge is equipped with a pump 7. The module 5 for decontamination solution receiving is equipped with a decontamination solution cleaning device made in the form of an ion-selective filter 8, and a device for pH correction of decontamination solution that comprises an alkaline substance supply pipe 9. In addition, the unit for electrochemical decontamination of metal radioactive waste, the module 5 for decontamination solution receiving and the decontamination solution preparation module 6 are equipped with pH measurement elements (not shown in the FIGURE). The unit for electrochemical decontamination of metal radioactive waste includes a cylindrical working container 10 with a lower conical part 11, a basket (not shown in the FIGURE) for the metal radioactive waste treated element (not shown in the FIGURE) installed inside the working container, the high-pressure water supply pipe 12 equipped with shut-off valves, as well as the pipe 4 for decontamination solution supply and discharge that are connected with the working container 10, as mentioned above. For the purposes of electrochemical deactivation, the unit for electrochemical decontamination of metal radioactive waste is provided with a DC source 13, the negative output of which is connected according to the cathode circuit, and a positive output is connected according to the anode circuit with a basket for the metal radioactive waste treated element. In addition, the unit for electrochemical decontamination of metal radioactive waste includes a mixing device 14 of the working container 10 that supplies compressed air to the cylindrical working container 10 connected with it, a precipitation discharge unit 15 located in the lower part of the working container 10. For the purposes of the decontamination solution condition monitoring, the unit for electrochemical decontamination of metal radioactive waste is equipped with an ionizing radiation power monitoring device (not shown in the FIGURE), a temperature monitoring device (not shown in the FIGURE) and a decontamination solution level monitoring device (not shown in the FIGURE) that are connected with the working container 10. In this embodiment, the module 5 for decontamination solution receiving includes a decontamination solution processing container 16 connected with the pipe 4 for decontamination solution supply and discharge equipped with an ion-selective filter 8, and an alkaline substance supply pipe 9 equipped with shut-off valves, the ionizing radiation power monitoring device (not shown in the FIGURE), the mixing device 17 supplying compressed air to the decontamination solution processing container 16 connected with it, the precipitation discharge device 18 located in the lower part of the decontamination solution processing container 16, and the decontamination solution level monitoring device (not shown in the FIGURE) located inside the decontamination solution processing container 16, that are connected with the decontamination solution processing container 16. The decontamination solution preparation module 6 includes a decontamination solution preparation container 19 connected with a water supply pipe 20 equipped with shut-off valves, an acidic substance supply pipe 21 equipped with shut-off valves, and a pipe 4 for decontamination solution supply and discharge, the mixing device 22 of the decontamination solution preparation container 19 connected with the decontamination solution preparation container 19 supplying compressed air to the connected decontamination solution preparation container 19, and a decontamination solution level monitoring device installed inside the decontamination solution preparation container 19 (not shown in the FIGURE). The ventilation module 3 includes a ventilation duct 2, a water separator 23 and a hydrogen post-combustion device 24 that are installed in it. The radioactive waste processing module 1 includes a metal radioactive waste degreasing unit containing a container 25 for metal radioactive waste degreasing connected with a degreasing solutions supply pipe (not shown in the FIGURE) equipped with shut-off valves, a high-pressure water supply pipe 12 equipped with shut-off valves, a steam supply pipe 26 equipped with shut-off valves, and an air supply pipe 27 equipped with shut-off valves, a pH measuring device (not shown in the FIGURE) and a precipitation discharge device 28 that are connected with a container 25 for metal radioactive waste degreasing. Also, the radioactive waste processing module 1 additionally includes an etching unit for non-metal coatings of metal radioactive waste containing a container 29 for etching of non-metal coatings of metal radioactive waste connected with an acidic substances supply pipe (not shown in the FIGURE) equipped with shut-off valves, a high-pressure water supply pipe 12 equipped with shut-off valves, and an air supply pipe 30 equipped with shut-off valves, a pH measuring device (not shown in the FIGURE) and a precipitation discharge device 31 that are connected with a container 29 for etching of non-metal coatings of metal radioactive waste. The radioactive waste processing module 1 includes a flushing unit connected by a high-pressure water supply pipe 12 equipped with shut-off valves with the unit for electrochemical decontamination of metal radioactive waste, the metal radioactive waste degreasing unit, and the etching unit for non-metal coatings of metal radioactive waste. The decontamination solution preparation module 6, the module 5 for decontamination solution receiving, the metal radioactive waste degreasing unit and the etching unit for non-metal coatings of metal radioactive waste are connected with the ventilation module 3 by a ventilation duct 2. The pipe 4 for decontamination solution supply and discharge is designed to transfer the decontamination solution from the decontamination solution preparation module 7 to the unit for electrochemical decontamination of metal radioactive waste and the module 5 for decontamination solution receiving. In addition, the pipe 4 for decontamination solution supply and discharge is designed to transfer the decontamination solution from the unit for electrochemical decontamination of metal radioactive waste to the module 5 for decontamination solution receiving and back. The plant for electrochemical decontamination of metal radioactive waste works as follows. The calculated amount of water is supplied to the decontamination solution preparation container 19 of the module 6 via the water supply pipe 20. Next, acidic substances (e.g., nitric acid) are supplied via the pipe 21, whilst the supply level of these substances is monitored by the decontamination solution level monitoring device. After this, the solution is mixed (bubbled) with compressed air supplied by the mixing device 22 with a simultaneous monitoring of the resulting solution pH level using a measurement element. Simultaneously with the bubbling from the decontamination solution preparation container 19, the air is removed by the ventilation duct 2 of the ventilation module 3. The prepared decontamination solution is supplied by pump 7 via the pipe 4 for decontamination solution supply and discharge into the working container 10 of the unit for electrochemical decontamination of metal radioactive waste of the module 1, while the element(s) of metal radioactive waste is(are) preliminarily installed in the basket of the working container 10. In the case of the presence of metal radioactive waste of non-metal deposits (contamination) on the element(s), the specified element(s) is(are) preliminarily subjected to chemical degreasing and etching described below. After filling the working container 10 with a decontamination solution to the calculated level, which is monitored by the decontamination solution level monitoring device, electric current is supplied using the source 13 according to the scheme where the negative output is connected according to the cathode circuit, and a positive output is connected according to the anode circuit with the basket for the metal radioactive waste treated element. Simultaneously with the filling of the working container 10 with a decontamination solution, the solution is mixed by compressed air (bubbling) supplied by a mixing device 14. During the decontamination process, the solution pH level, the temperature and the power of ionizing radiation are monitored using measuring elements. The air containing hydrogen and water vapour is removed by a ventilation channel 2 of a module 3 through a water separator 37 and a hydrogen post-combustion device 38. After the decontamination process is completed, the decontamination solution is completely discharged by a pump 7 through a pipe 4 for decontamination solution supply and discharge into a decontamination solution processing container 16 of a module 5 through an ion-selective filter 8. At the same time, the ion-selective filter 8 installed in the pipe 4 for decontamination solution supply and discharge is designed to clean the contaminated spent decontamination solution (electrolyte) flowing from the container 10 from the Cz137, Co60 radionuclides (isotopes of technogenic origin) by precipitating them in the inorganic selective sorbent phase. The operating time until the absorption capacity of the filter 8 has been exhausted and the transport and technological operation for its transfer to the place of organized long-term storage are determined during the development of the technology and design binding to a specific facility. The filling of the filter 8 cavity with inorganic selective sorbent to the design capacity of its body is performed at the place of use. Project sorbents have selective ability to radionuclides Cz137 and Co60. After discharge of the decontamination solution from the container 10, the treated (deactivated) element(s) of metal radioactive waste is(are) subjected to hydrodynamic processing using a high-pressure water supply pipe 12, while this processing can be performed outside the container 10 in the absence of a high pressure water supply pipe 12 in the plant for electrochemical decontamination of metal radioactive waste. The precipitations remaining in the working container 10 are discharged using the precipitation discharge unit 15 located in the conical part 11 of the working container 10. The module 5 for decontamination solution receiving is designed to neutralize contaminated spent decontamination solution containing nitric acid with sodium hydroxide and calcium. The solution pH and the ionizing radiation power are monitored using measuring elements in the decontamination solution processing container 16 of the module 5. Based on the obtained measurements, sodium and calcium hydroxide is supplied using an alkaline substance supply pipe 9 to a container 16 with a simultaneous mixing (bubbling) of the solution by compressed air supplied by a mixing device 17. Next, the solution pH and the power of ionizing radiation are repeatedly monitored. On the basis of the data obtained, the decision is taken to re-process the spent decontamination solution or about the readiness of the purified decontamination solution to be supplied to the working container 10 of the electrochemical decontamination unit. The container 16 is cleaned from the remnants of the radionuclides of the corrosion group by precipitating them in the sludge after opening the shut-off valve 18. During the cleaning of the spent decontamination solution in the container 16, the air is removed by the ventilation duct 2 of the ventilation module 3. In the case of a shortage of purified decontamination solution in the container 16, the level of which is monitored using a decontamination solution level monitoring device, the previously prepared decontamination solution is topped up from a decontamination solution preparation container 19 of a module 6 by supplying decontamination solution with a pump 7 via pipe 4 for decontamination solution supply and discharge in the container 16 of a module 5 for decontamination solution receiving. The prepared calculated amount of the purified decontamination solution is supplied from the container 16 of the module 5 for decontamination solution receiving using a pump 7 via the pipe 25 for decontamination solution supply and discharge to the working container 10 of the unit for electrochemical decontamination of metal radioactive waste of a module 1. At the same time, the same treated element(s) of metal radioactive waste is(are) installed in the basket of the working container 13, if it is necessary to repeat its decontamination, or a new treated element(s) of metal radioactive waste is present, the decontamination procedure is repeated as described above. As mentioned above, in the case of the presence of metal radioactive waste of non-metal deposits (contamination) on the element(s), prior to the electrochemical decontamination, chemical degreasing and etching are carried out. For degreasing, the treated element(s) of metal radioactive waste is(are) installed in the container 25 of the metal radioactive waste degreasing unit of the radioactive waste processing module 1. Next, highly alkaline degreasing solutions, in particular, soda ash, and heating steam, are supplied via the degreasing solution supply pipe (not shown in the FIGURE) and the steam supply pipe 26 to the container 25. At the same time, pressurized air is supplied to the container 25 for mixing of the above substances, which ensures effective degreasing of the surface of the treated element(s) of metal radioactive waste by an air supply pipe 27. After the degreasing process, the surface of the treated element(s) of metal radioactive waste is(are) subjected to hydrodynamic processing using a high-pressure water supply pipe 12, while this processing can be performed outside the container 25 in the absence of a high pressure water supply pipe 12 in the plant for electrochemical decontamination of metal radioactive waste. During the degreasing, the air is removed from the container 25 by the ventilation duct 2 of the ventilation module 3. After the hydrodynamic processing, sediments, mainly consisting of non-metal deposits (contamination), are unloaded from the container 25 by a device 28. After the hydrodynamic processing, the element(s) of metal radioactive waste is(are) installed in the container 29 of the etching unit for non-metal coatings of metal radioactive waste of the radioactive waste processing module 1. Next, acidic substances, mainly sulfuric or hydrochloric or phosphoric acids, are supplied via the acidic substance supply pipe to the container 29. At the same time, pressurized air is supplied to the container 29 for mixing of the acidic substances, which ensures effective etching of the surface of the element(s) of metal radioactive waste by an air supply pipe 30. After the etching process, the surface of the treated element(s) of metal radioactive waste is(are) subjected to hydrodynamic processing using a high-pressure water supply pipe 12, while this processing can be performed outside the container 29 in the absence of a high pressure water supply pipe 12 in the plant for electrochemical decontamination of metal radioactive waste. During the etching, the air is removed from the container 29 by the ventilation duct 2 of the ventilation module 3. After the hydrodynamic processing, sediments, mainly consisting of non-metal deposits (contamination), are unloaded from the container 29 by a device 31. |
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abstract | A scintillator panel for converting radiation into scintillation light, the scintillator panel includes a substrate having a front surface and a back surface, and formed with a plurality of convex portions projecting from the front surface in a predetermined direction toward the front surface from the back surface and a concave portion defined by the convex portions, a plurality of first scintillator sections formed on the respective convex portions of the substrate, and a second scintillator section formed on the bottom surface of the concave portion of the substrate, and the first scintillator section has a first portion extending along the predetermined direction from an upper surface of the convex portion and a second portion extending along the predetermined direction from side surfaces of the convex portion so as to contact with the first portion, the first and second portions are composed of a plurality of columnar crystals of a scintillator material, the first scintillator sections are separated from one another, and the second scintillator section is in contact with the second portion. |
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abstract | An ultraviolet device used for flooding an air duct of an air ventilation system with ultraviolet light comprising a mounting portion, the mounting portion that is mountable to an air duct, at least one mounting bracket which is interchangeably mountable to the mounting portion and at least one ultraviolet light lamp, the lamp is mountable to the mounting bracket wherein the angle at which the lamp mounts to said mounting bracket may be configured to maximize the coverage of ultraviolet light within the air duct. |
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description | This application claims priority under 35 U.S.C. §119(e) to U.S. patent application Ser. No. 61/614,508 entitled “SINTER BONDED CONTAINMENT TUBE,” by Banach et al., filed Mar. 22, 2012, which is assigned to the current assignee hereof and incorporated herein by reference in its entirety. This disclosure, in general, relates to containment tubes, and more particularly to sealed containment tubes comprising silicon carbide. A sintered ceramic sealed tube has first and second ends and an inner bore extending along at least a portion of its axial length between the first and second ends, the first end having a plug residing in the inner bore to close the first end, the second end having a distal wall to close the inner bore at the second end. The ceramic tube, or the plug, or both, may comprise silicon carbide, and in certain embodiments comprise principally silicon carbide, such that silicon carbide is the majority compositional species of the tube. The ceramic sealed tube includes a sinter bond between at the tube and at least the distal wall or the plug, such that the sinter bond forms a hermetic seal, or interference bond, that includes no bond materials. The use of the same reference symbols in different drawings indicates similar or identical items. Embodiments of the present invention are generally drawn to containment tubes and methods for forming containment tubes. In one embodiment, a containment tube includes a sealed tube comprising silicon carbide, the sealed tube having a generally constant diameter along its axial length and containing a radioactive material. A “generally constant diameter” means that the outer diameter of the tube does not vary considerably from a nominal or average diameter value. According to one embodiment, any measureable diameter variances do not exceed 15% of the nominal diameter value, such as not greater than 10%, not great than 5%, not greater than 4%, not greater than 3%, not greater than 2%, or not greater than 1%. In one embodiment, to the naked eye, the containment tube appears to be uniform and rectilinear. In particular, in one embodiment, a sintered ceramic sealed tube has first and second ends and an inner bore extending along at least a portion of its axial length between the first and second ends, the first end having a plug residing in the inner bore to close the inner bore at the first end, the second end having a distal wall to close the second end. The ceramic tube, or the plug, or both, may comprise silicon carbide, and in certain embodiments comprise principally silicon carbide, such that silicon carbide is the majority compositional species of the tube, typically greater than at least about 70 wt %, such as greater than at least about 80 wt %, or greater than at least about 90 wt %. In another embodiment, the tube may comprise silicon carbide in an amount greater than at least about 91 wt %, at least about 95%, at least about 99%, at least about 95%, at least about 99.85 wt %. One particular form of silicon carbide is used according to certain embodiments, known as HEXOLOY®-brand silicon carbide (manufactured by Saint-Gobain Advanced Ceramics Corporation of Worcester, Mass., USA), described in U.S. Pat. No. 4,179,299 incorporated herein by reference. Suitable silicon carbides generally contain silicon carbide in an amount greater than at least about 91 wt %, such as greater than at least about 99.85 wt %, up to about 5.0 wt % carbonized organic material, from at least about 0.15 wt % to not greater than about 3.0 wt % boron, and up to about 1.0 wt % additional carbon. The “carbonized organic material” is free carbon or uncombined carbon produced in situ by the carbonization of the organic material used as a raw material in the process of forming the ceramic tube. The carbonizable organic materials that can be used in forming the ceramic tube are well known in the art, and include but are not limited to phenolic resin, coal tar pitch, polyphenylene, or polymethylphenylene, and the like. Sintered ceramic bodies of silicon carbide according to an embodiment may be characterized by a predominantly equiaxed microstructure, meaning the presence of grains having an aspect ratio of less than 3:1 (i.e., the ratio of the maximum dimension of the grains of the crystal microstructure to the minimum dimension of the grains of the crystal microstructure is less than 3:1). Moreover, the silicon carbide comprises at least about 95 wt %, such as at least about 99 wt % alpha-phase, non-cubic crystalline silicon carbide. The density of silicon carbide according to an embodiment is at least about 2.40 g/cm3, such as at least about 2.90 g/cm3, or at least about 3.05 g/cm3. A better understanding of the embodiments of the present invention may be better had with reference to the drawings. In particular, in connection with FIG. 1, an embodiment of a containment tube is illustrated. As shown, the containment tube 10 has a generally elongate body, which may be described or quantified in terms of aspect ratio, which is the ratio of length to outside diameter. With respect to the relationship of length (L) to outer diameter (DO) referred to herein as aspect ratio, generally the tube will have an aspect ratio of not less than about 10:1, such as not less than about 20:1, such as not less than about 30:1, or not less than about 40:1. Typically, the aspect ratio is limited, as extended length tubes are difficult to handle and fully sinter. Consequently, aspect ratios typically do not exceed 300:1. The containment tube 10 includes a first end 14, and a second end 12. Along the inner bore 13 of the containment tube, is provided a plug 16 at the first end, closing the first end hermetically. In an embodiment, the plug 16 closes the first end and provides a hermitic seal by way of a sinter bond, or interference bond, between the plug 16 and the first end 14 of the tube 10. The second end has a distal wall 15 which, in this particular embodiment, is integrated into the outer wall forming the outside diameter of the tube. In an embodiment, the distal wall 15 is integrated into the outer wall by a sinter bond, or interference bond. Radioactive material 18 is disposed within the tube, generally remote from first end 14 and plug 16. Turning to FIG. 2, another embodiment of a containment tube is illustrated. Containment tube 20 includes a first end 21, a second end 22, a bore 23, and first and second plugs 24 and 25 respectively provided to close the bore at first and second ends 21, 22, respectively. That should be clear, the containment tube 20 differs from containment tube 10 in that containment tube 20 has a dual-plug structure. Accordingly, at least one of the first end 21 and the second end 22 of the containment tube 20 includes a sinter bond, or interference bond, with the respective first or second plugs 24 and 25. According to a particular feature, containment tubes according to embodiments herein may be formed through a multi-step sintering process. Turning to FIG. 3, a green containment tube 30 is shown, having an inner bore 33, an outer wall 34, first and second ends 31, 32 respectively, the second end 32 having a distal wall 35 closing the bore at the second end 32. The green tube 30 may be formed of any one of known manufacturing techniques. Although various forming techniques can be utilized for fabrication of the tube, such as slip casting, isopressing, machining of large stock materials, and other forming techniques, extrusion may be used according to particular embodiments. Extrusion represents a cost-effective and desirable fabrication approach for making multiple articles requiring tubes of varying lengths and diameters. Sintered ceramic bodies of silicon carbide according to an embodiment may be characterized by the amount the bodies shrink from a green state to a fully sintered state. For example, green ceramic bodies of silicon carbide according to an embodiment may shrink more than about 10% from their original size, more than about 12%, more than about 15%, more than about 17%, less than about 25%, less than about 20%, less than about 17%, less than about 15% upon being fully sintered. In a particular embodiment, a green ceramic body of silicon carbide may shrink approximately 17% from its original size upon being fully sintered. When combining a pre-sintered first component, such as a plug or distal wall, with a green second component, such as an un-sintered portion of a tube, that circumvents the pre-sintered first component, the shrinkage relationship, and the amount of interference bond, can be expressed as follows.IDt,FS=ODp−Δ, where IDt,FS is the inside diameter (ID) of a fully sintered tube, ODp is the outside diameter (OD) of the pre-sintered plug, and Δ is the intereference (tube undersizement). For example, a pre-sintered plug has a bond surface, or outside diameter, of 2.0″ (i.e. ODt=2.0). An interference bond of 5% (i.e. Δ=5%) of a second body, such as the tube, requires a fully sintered tube ID (IDt,FS) to be 0.10″ less than the ODp (i.e. 2.0*5%=0.10), or 1.90″ (i.e. IDt,FS=ODp−Δ, or 1.90″=2.0″−0.10″). Thus, to attain a 5% interference of a fully sintered tube on the pre-sintered plug, the green portion of the tube (i.e. the un-sintered portion of the tube) will be made to have a theoretically fully sintered inner diameter (if it were sintered by itself) of 1.90″. Further, the ID of the green second component (i.e. the un-sintered portion of the tube) can be expressed as follows.IDt,FS/(1−Rs)=IDt, where IDt is the inner diameter of the green second component, or un-sintered portion of the tube, and Rs is the shrinkage rate of the second component (expressed as a decimal). Thus, in accordance with the example given above, and assuming the shrinkage rate of the second component is 17.0%, the inner diameter of the green portion of the tube (IDt) can be calculated as 1.9÷(1−0.170)=2.289″. Following appropriate shape formation (i.e. forming of the green ceramic tube), the green tube 30 may be subjected to a machining operation during which the outer surface of the ceramic tube is machined prior to pre-sintering. Stated alternatively, this machining step is carried out in the green state, where the tube is in a state that allows easier material removal than in the sintered state. Moreover, the machining may be effective to reduce or even completely remove dimensional (out-of-roundness) or surface irregularities of the green tube. For example, in the case of extrusion, the green tube may have characteristic score lines extending partially or wholly along the entire length of the tube. Those score lines can inhibit the formation of a strong interfacial sinter bond, as well as a hermetic seal. In the case of other formation technologies, machining may still be desirable. For example, in the case of isopressing or molding, characteristic imperfections may be left behind on the green tube, such as a flashing. Both the surface cleaning and machining steps may be carried out through mechanical abrasion processes. Mechanical abrasion can include machining using a free abrasive (e.g., an abrasive slurry), a coated abrasive, or a fixed abrasive. The species of abrasive product is chosen to prevent unwanted chemical interaction with or foreign deposits on the tube, while also providing adequate material removal rates. Generally speaking in the case of silicon carbide, abrasive materials such as alumina are avoided, and materials such as silicon carbide and superabrasives, notably including cubic boron nitride (CBN) and diamond, are utilized. In the green state, machining may be carried out with silicon carbide and in the sintered state, surface cleaning may be done with silicon carbide or a superabrasive species. In practice, embodiments have made use of coated abrasives, such as a silicon carbide, CBN, or diamond abrasive coated on a closed looped belt, mounted to a belt sander. While the cleaning steps above set forth in connection with a tube, particularly the outer surface of the tube, the foregoing cleaning operations can be carried out with respect to an inner surface of the tube and particularly the plugs. According to one particular feature, processing may continue with partial sintering of the tube. As shown in FIG. 3, partial sintering 36 is carried out with respect to a portion of the tube including a middle portion extending to the second end, leaving the first end unheated, or only partially sintered such that the second end does not shrink to its final dimensions. Upon completion of partial sintering 36, the tube is in a hybrid state, a central portion of the tube extending to the second end being fully sintered and shrunk to its final dimensions, the first end being still in the green state or only partially sintered. This hybrid state can be better seen in connection with FIG. 4, in which containment tube 40 shows sintered portion 42, and partially or un-sintered portion 44 extending to the first end 46. As can be seen, the first end 46 has a larger inside diameter and outside diameter relative to the second end 48 which has already undergone densification and shrinkage. Typically, the differences in outside diameter from the first end to the second end are on the order of 1-20%, sometimes larger. Processing continues with incorporation with radioactive material 41 into the inner bore 43 of the tube, and placement of plug 45 in the bore at the first end 46. Typically, the plug used to seal one or both ends of the containment tube structure is formed of the same material as the tube, in accordance with the above description. Upon completion of forming a green ceramic plug, the plug proceeds to a pre-sintering step to form a sintered plug. Pre-sintering can be carried out in any one of known furnaces, including continuous furnaces or a batch furnace that translate the work piece (herein, the plug) through the furnace at a constant or variable rate. Pre-sintering is generally carried out at a temperature above 2000° C., such as above 2050° C., but generally below 2400° C., such as below 2300° C., such as below 2250° C. A suitable target range for sintering the green ceramic plug in the case of silicon carbide can lie within a range of 2100-2200° C. Sintering times can vary, and are largely dependent on the thermal mass of the plug. However, typically sintering times range from 15 minutes to 10 hours, such as not less than about 30 minutes, such as not less than about 1 hour, such as not less than about 1.5 hours. While large, high mass plugs may require extended sintering times, typically sintering times do not exceed 30 hours, such as not great than 20 hours, such as not greater than 10 hours. Accordingly, the machining operations applied the tube can also be applied to the outside of the plug. After the sintering step is completed, at least a portion of an outer surface of the sintered plug is subjected to surface cleaning. Typically, at least the portion of the plug that will contact the base component will be subjected to surface cleaning. In this respect, it has been found that the outer surface of the plug can carry contaminates, such as contaminates that are deposited during the sintering process, or which form as a consequence of the sintering process and changes in the crystallographic and compositional structure of the plug. For example, binders within the composition may burn-out, leaving behind a carbonaceous residue on the outer surface of the plug. That carbonaceous residue, generally in the form of free carbon, can negatively impact the quality of bond between the plug and the base component, inhibiting a hermetic seal. Thereafter, sintering 47 (i.e. co-sintering the first end 46 and the plug 45) takes place to complete the sintering of the partially or un-sintered portion 44, shrinking it to its final dimensions, forming a completed containment tube having a generally constant outside diameter, and resembling the structure shown in FIG. 1, and providing a sinter bond between the first end 46 and the plug 45. The sinter bond between the first end and the plug is defined as an interface bond, or an interference bond, and includes no bond materials. The green silicon carbide material of the un-sintered portion 44 of the first end 46 shrinks to some degree upon sintering, and the quality of the interference bond is at least in part due to selecting a size of the green, un-sintered second portion 44. As discussed above, the quality of the interference bond is also attributable to preparing the surface of the pre-sintered silicon carbide component to remove contaminants from its surface. The interface bond has at least one of the following performance characteristics: a Shear Strength not less than about 25 MPa, a Nitrogen Seal Performance of not greater than 10%, a Helium Seal Performance of not greater than 10%, and/or a Vacuum Seal Performance of not greater than 10%. In one embodiment, the interface between the first and the second component exhibits a Shear Strength not less than about 25 MPa, not less than about 40 MPa, not less than about 50 MPa, not less than about 75 MPa, not less than about 100 MPa, not less than about 120 MPa, not less than about 140 MPa, not less than about 170 MPa, or not less than about 200 MPa. In one embodiment, the interface between the first and the second component exhibits a Shear Strength not greater than about 1000 MPa, such as not greater than about 700 MPa, not greater than about 500 MPa, or not greater than about 300 MPa. As used herein, reference to Shear Strength as a particular Shear Strength value is measured by testing a sample having standardized dimensions under load. In particular, the Shear Strength is measured by preparing and testing a standardized sample as follows. The sample is prepared from a ceramic tube and a ceramic ring, each having a length of 76.2 mm. The ceramic tube has an outer diameter (ODt) of 14 mm and an inner diameter (IDt) of 11 mm. The ceramic ring has an outer diameter (ODr) of 20 mm, and an inner diameter (IDr) of 14 mm. The ceramic ring is placed around the ceramic tube so that the ends of each are flush, and the tube-ring assembly is then co-sintered. After cooling, a cross-sectional center segment is sliced from the sintered assembly and thickness grinded to a final thickness (t) of 3 mm. The center segment comprises an inner ring sliced from the ceramic tube and an outer ring sliced from the ceramic ring. The area of contact between the inner and outer rings represents the total bond area (Ab), and is calculated according to the following formula:Ab=π·ODt·t (Formula I) The Shear Strength of the center segment sample is tested at room temperature using an Instron 8562 using a 100 kN load cell at a speed of 0.05 mm/min, which applies equal but opposing force to the inner and outer rings, respectively. The magnitude of the applied force is gradually increased until the rings break apart. The force (F) required to break the rings apart is measured in Newtons. The Shear Strength (t) value is obtained according to the following formula:τ=F·Ab·106 (Formula II) It should be understood that ceramic articles as described herein can be a wide variety of dimensions and overall sizes, but the Shear Strength values are based on a standardized geometry and testing approach as described above. Consequently, validating the Shear Strength of a sample having differing dimensions larger or smaller than the standardized sample described above requires the fabrication of a standardized sample under identical compositional and processing conditions to that of the sample having differing dimensions. A Nitrogen Seal Performance is determined according to a nitrogen seal performance test, wherein nitrogen is applied at an interface of a seal at a given initial positive pressure, and pressure loss is measured by a pressure gauge. Nitrogen Seal Performance is the percent pressure drop occurring across the seal interface over a period of 2 hours at an applied gauge pressure, such as 200 psi. Embodiments herein achieve a Nitrogen seal performance of not greater than 10%, not greater than 9%, not greater than 8%, not greater than 7%, not greater than 6%, not greater than 5%, not greater than 4%, not greater than 3%, not greater than 2%, not greater than 1.9%, not greater than 1.8%, not greater than 1.7%, not greater than 1.6%, not greater than 1.5%, not greater than 1.4%, not greater than 1.3%, not greater than 1.2%, not greater than 1.1%, not greater than 1.0%, not greater than 0.9%, not greater than 0.8%, not greater than 0.7%, not greater than 0.6%, not greater than 0.5%, not greater than 0.4%, not greater than 0.3%, not greater than 0.2%, or not greater than 0.1% of an initial pressure differential of 200 PSI (gauge pressure). A Helium Seal Performance is determined according to a helium seal performance test, wherein helium is applied at an interface of a seal at a given initial positive pressure and pressure loss is measured by a pressure gauge. Helium Seal Performance is achieved if the pressure drop occurring across the seal interface over a period of 2 hours is not greater than 10%, not greater than 9%, not greater than 8%, not greater than 7%, not greater than 6%, not greater than 5%, not greater than 4%, not greater than 3%, not greater than 2%, not greater than 1.9%, not greater than 1.8%, not greater than 1.7%, not greater than 1.6%, not greater than 1.5%, not greater than 1.4%, not greater than 1.3%, not greater than 1.2%, not greater than 1.1%, not greater than 1.0%, not greater than 0.9%, not greater than 0.8%, not greater than 0.7%, not greater than 0.6%, not greater than 0.5%, not greater than 0.4%, not greater than 0.3%, not greater than 0.2%, or not greater than 0.1% of an initial pressure differential of 87 PSI (gauge pressure), an initial pressure differential of about 200 psi (about 13.8 bar), or an initial pressure differential of about 6 barg (bar gauge). A Vacuum Seal Performance is determined according to a vacuum seal performance test. In the vacuum seal performance test, a vacuum is applied to a seal. The nitrogen gas atmosphere inside the tube is then lowered from 1 ATM (760 torr) to a pressure of 10 torr thereby having a pressure differential of 750 torr. Vacuum Seal Performance is achieved if the gain inside the tube occurring across the seal interface over a period of 2 hours is not greater than 10%, not greater than 9%, not greater than 8%, not greater than 7%, not greater than 6%, not greater than 5%, not greater than 4%, not greater than 3%, not greater than 2%, not greater than 1.9%, not greater than 1.8%, not greater than 1.7%, not greater than 1.6%, not greater than 1.5%, not greater than 1.4%, not greater than 1.3%, not greater than 1.2%, not greater than 1.1%, not greater than 1.0%, not greater than 0.9%, not greater than 0.8%, not greater than 0.7%, not greater than 0.6%, not greater than 0.5%, not greater than 0.4%, not greater than 0.3%, not greater than 0.2%, or not greater than 0.1% of the pressure differential (750 torr). In each of the seal performance tests, the bond or interface is subjected to the above-described pressure differential. Depending on the geometry of the part, an inner volume is pressurized or evacuated, and holes plugged. In a case of an external seal, such as in the case of a flange on a tube, an end-cap is positioned to cover the flange and exposed bore of the tube, the cap being offset from the bore to allow fluid communication (and hence pressure/vacuum) extending radially up to the bond region. Caps/plugs can have varying geometries to fit the part undergoing test, and can be sealed with a vacuum grease to ensure a pressure tight, hermetic seal. |
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claims | 1. A closure for shielding, and selectively providing access to, the targeting assembly of a particle accelerator, the particle accelerator including a housing defining an opening for accessing the targeting assembly, the particle accelerator being surrounded by an outer shielded enclosure providing selective access to the particle accelerator, said closure being adapted to be mounted on said housing and comprising at least a first door for selectively covering the opening in the housing of the particle accelerator, and said closure including a door mounting assembly for mounting said first door on the housing of the particle accelerator, whereby said first door of said closure selectively covers the opening in the housing of the particle accelerator when access to the particle accelerator through the outer shielded enclosure is provided. 2. The closure of claim 1 wherein said first door includes copper radiation shielding. 3. The closure of claim 1 wherein said door mounting assembly includes at least a first hinge assembly to facilitate pivotally mounting said first door on the housing of the particle accelerator. 4. The closure of claim 1 wherein said door mounting assembly includes a frame for being mounted on the housing of the particle accelerator and received about the opening in the housing of the particle accelerator and for supporting said door. 5. The closure of claim 4 wherein said door mounting assembly includes at least a first hinge assembly for pivotally mounting said door to said frame. 6. The closure of claim 4 wherein said frame and said door include copper shielding material. 7. The closure of claim 4 wherein said frame and said door are fabricated substantially of copper. 8. The closure of claim 1 wherein said door mounting assembly includes a frame for being received about the opening in the housing of the particle accelerator, said frame including a sill member, a header member, and first and second jamb members, said door mounting assembly also including at least a first hinge assembly for pivotally mounting said door on said frame, whereby said first door is movable from a closed position to an open position. 9. The closure of claim 8 wherein said closure further comprises a second door, and said door mounting assembly includes a second hinge assembly for pivotally mounting said second door on said frame, whereby said second door is movable from a closed position to an open position. 10. The closure of claim 9 wherein each said first and second door is substantially rectangular and defines outboard and inboard edges, and upper and lower edges, and wherein said each said first and second jamb member defines a front surface, said outboard edge of said first door being pivotally secured to said first sill member with said first hinge assembly such that said first door covers said front surface of said first jamb member when said first door is in said closed position, and said outboard edge of said second door being pivotally secured to said second sill member with said second hinge assembly such that said second door covers said front surface of said second jamb member when said second door is in said closed position. 11. The closure of claim 10 wherein said sill member of said frame defines a first rabbet along an upper forward edge of said sill member for receiving said lower edges of said first and second doors when said first and second doors are in said closed position, and wherein said header member of said frame defines a second rabbet along a lower forward edge of said header member for receiving said upper edges of said first and second doors when said first and second doors are in said closed position. 12. The closure of claim 11 wherein said first door defines a third rabbet along the inside of said inboard edge of said first door, and wherein said second door defines a fourth rabbet along the outside of said inboard edge of said second door, whereby said inboard edges of said first and second doors overlap when said first and second doors are in said closed position. 13. The closure of claim 12 wherein said first and second doors and said frame are fabricated substantially of copper. 14. A closure for shielding, and selectively providing access to, the targeting assembly of a particle accelerator, the particle accelerator including a housing defining an opening for accessing the targeting assembly, the particle accelerator being surrounded by an outer shielded enclosure providing selective access to the particle accelerator, said closure comprising:first and second doors for selectively covering the opening in the housing of the particle accelerator, each said first and second door being movable from a closed position whereby the targeting assembly is shielded to an open position, whereby access to the targeting assembly is provided, anda door mounting assembly for mounting said first and second doors on the housing of the particle accelerator, said door mounting assembly including a frame for being secured about the opening in the particle accelerator accessing the targeting assembly, said door mounting assembly also including a first hinge assembly for pivotally securing said first door to said frame and a second hinge assembly for pivotally securing said second door to said frame, whereby said first and second doors of said closure selectively cover, and reduce radiation emissions from, the opening in the housing of the particle accelerator and the targeting assembly therein when access to the particle accelerator through the outer shielded enclosure is provided. 15. The closure of claim 14 wherein said first and second doors are fabricated substantially of copper. 16. The closure of claim 15 wherein said frame is fabricated substantially of copper. 17. The closure of claim 14 wherein said frame includes a sill member, a header member, and first and second jamb members. 18. The closure of claim 17 wherein each said first and second door is substantially rectangular and defines outboard and inboard edges, and upper and lower edges, and wherein said each said first and second jamb member defines a front surface, said outboard edge of said first door being pivotally secured to said first sill member with said first hinge assembly such that said first door covers said front surface of said first jamb member when said first door is in said closed position, and said outboard edge of said second door being pivotally secured to said second sill member with said second hinge assembly such that said second door covers said front surface of said second jamb member when said second door is in said closed position. 19. The closure of claim 18 wherein said sill member of said frame defines a first rabbet along an upper forward edge of said sill member for receiving said lower edges of said first and second doors when said first and second doors are in said closed position, and wherein said header member of said frame defines a second rabbet along a lower forward edge of said header member for receiving said upper edges of said first and second doors when said first and second doors are in said closed position. 20. The closure of claim 19 wherein said first door defines a third rabbet along the inside of said inboard edge of said first door, and wherein said second door defines a forth rabbet along the outside of said inboard edge of said second door, whereby said inboard edges of said first and second doors overlap when said first and second doors are in said closed position. 21. The closure of claim 20 wherein said closure further comprises a locking mechanism for securing said first and second doors in said closed position. 22. The closure of claim 21 wherein said locking mechanism includes a first and second securing pins, said first securing pin being releasably received through a hole in said header member, and releasably received in a hole provided in said first door, and said second securing pin being releasably received through a further hole in said header member, and releasably received in a hole provided in said second door. 23. A closure for shielding, and selectively providing access to, the targeting assembly of a particle accelerator, the particle accelerator including a housing defining an opening for accessing the targeting assembly, the particle accelerator being surrounded by a shielded enclosure providing selective access to the particle accelerator, said closure comprising:first and second doors for selectively covering the opening in the housing of the particle accelerator, each said first and second door being fabricated substantially of copper and being movable from a closed position whereby the targeting assembly is shielded to an open position whereby access to the targeting assembly is provided, anda door mounting assembly for mounting said first and second doors on the housing of the particle accelerator, said door mounting assembly including a frame for being secured about the opening in the particle accelerator accessing the targeting assembly, said frame being fabricated substantially of copper, said door mounting assembly also including a first hinge assembly for pivotally securing said first door to said frame and a second hinge assembly for pivotally securing said second door to said frame, whereby said first and second doors of said closure selectively cover, and reduce radiation emissions from, the opening in the housing of the particle accelerator and the targeting assembly therein when access to the particle accelerator is provided through the shielded enclosure. 24. The closure of claim 23 wherein said first door defines an interior surface which is contoured to closely receive components of the targeting assembly of the particle accelerator. 25. The closure of claim 23 wherein each said first and second door is substantially rectangular and defines outboard and inboard edges, and upper and lower edges, and wherein said each said first and second jamb member defines a front surface, said outboard edge of said first door being pivotally secured to said first sill member with said first hinge assembly such that said first door covers said front surface of said first jamb member when said first door is in said closed position, and said outboard edge of said second door being pivotally secured to said second sill member with said second hinge assembly such that said second door covers said front surface of said second jamb member when said second door is in said closed position. 26. The closure of claim 25 wherein said sill member of said frame defines a first rabbet along an upper forward edge of said sill member for receiving said lower edges of said first and second doors when said first and second doors are in said closed position, and wherein said header member of said frame defines a second rabbet along a lower forward edge of said header member for receiving said upper edges of said first and second doors when said first and second doors are in said closed position. 27. The closure of claim 26 wherein said first door defines a third rabbet along the inside of said inboard edge of said first door, and wherein said second door defines a forth rabbet along the outside of said inboard edge of said second door, whereby said inboard edges of said first and second doors overlap when said first and second doors are in said closed position. 28. A closure for shielding, and selectively providing access to, the targeting assembly of a particle accelerator, the particle accelerator including a housing defining an opening for accessing the targeting assembly, the particle accelerator being surrounded by an outer shielded enclosure providing selective access to the particle accelerator, said closure being adapted to be mounted on said housing and comprising at least a first door for selectively covering the opening in the housing of the particle accelerator, and said closure including a door mounting assembly for mounting said first door on the housing of the particle accelerator, whereby said first door of said closure selectively covers the opening in the housing of the particle accelerator when access to the particle accelerator through the outer shielded enclosure is provided, said door defining an interior surface having a contour adapted to be closely received over at least one component of the targeting assembly of the particle accelerator. 29. The closure of claim 28 wherein said first door includes copper radiation shielding. 30. The closure of claim 28 wherein said door mounting assembly includes at least a first hinge assembly to facilitate pivotally mounting said first door on the housing of the particle accelerator. 31. The closure of claim 28 wherein said door mounting assembly includes a frame for being mounted on the housing of the particle accelerator and received about the opening in the housing of the particle accelerator and for supporting said door. 32. The closure of claim 31 wherein said door mounting assembly includes at least a first hinge assembly for pivotally mounting said door to said frame. 33. The closure of claim 31 wherein said frame and said door include copper shielding material. 34. The closure of claim 31 wherein said frame and said door are fabricated substantially of copper. 35. The closure of claim 28 wherein said door mounting assembly includes a frame for being received about the opening in the housing of the particle accelerator, said frame including a sill member, a header member, and first and second jamb members, said door mounting assembly also including at least a first hinge assembly for pivotally mounting said door on said frame, whereby said first door is movable from a closed position to an open position. 36. The closure of claim 35 wherein said closure further comprises a second door, and said door mounting assembly includes a second hinge assembly for pivotally mounting said second door on said frame, whereby said second door is movable from a closed position to an open position. 37. The closure of claim 36 wherein each said first and second door is substantially rectangular and defines outboard and inboard edges, and upper and lower edges, and wherein said each said first and second jamb member defines a front surface, said outboard edge of said first door being pivotally secured to said first sill member with said first hinge assembly such that said first door covers said front surface of said first jamb member when said first door is in said closed position, and said outboard edge of said second door being pivotally secured to said second sill member with said second hinge assembly such that said second door covers said front surface of said second jamb member when said second door is in said closed position. 38. The closure of claim 37 wherein said sill member of said frame defines a first rabbet along an upper forward edge of said sill member for receiving said lower edges of said first and second doors when said first and second doors are in said closed position, and wherein said header member of said frame defines a second rabbet along a lower forward edge of said header member for receiving said upper edges of said first and second doors when said first and second doors are in said closed position. 39. The closure of claim 38 wherein said first door defines a third rabbet along the inside of said inboard edge of said first door, and wherein said second door defines a fourth rabbet along the outside of said inboard edge of said second door, whereby said inboard edges of said first and second doors overlap when said first and second doors are in said closed position. 40. The closure of claim 39 wherein said first and second doors and said frame are fabricated substantially of copper. |
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abstract | A lithographic projection apparatus has a discharge plasma radiation source that is contained in a vacuum chamber. The radiation source is to generate a beam of EUV radiation. A chamber wall of the vacuum chamber incorporates a channel structure comprising adjacent narrow channels separated by walls that are substantially parallel to a propagation direction of the radiation generated so as to pass the radiation from the vacuum chamber through the structure to another subsequent vacuum chamber. In the subsequent vacuum chamber, a much higher vacuum level (lower pressure) can be maintained than is present in the vacuum chamber of the radiation source. |
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abstract | A scintillator panel includes a substrate made of an organic material, a barrier layer formed on the substrate and including thallium iodide as a main component, and a scintillator layer formed on the barrier layer and including cesium iodide as a main component. According to this scintillator panel, moisture resistance can be improved by providing the barrier layer between the substrate and the scintillator layer. |
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abstract | A focused ion beam (FIB) system that can automatically set processing and scanning conditions under which a specimen is processed includes an arithmetic unit for selecting optical conditions for condenser lenses, multiple variable apertures, beam-deflecting electrodes, and an objective lens based on data entered into the input device. The arithmetic unit automatically calculates the processing and scanning conditions under which the specimen is processed by the focused ion beam, according to the selected-optical conditions. The system further includes a setting condition data output portion for outputting data based on the optical conditions and processing and scanning conditions selected and calculated by the arithmetic unit. The system further includes a FIB driver portion for driving the condenser lenses, beam-blanking electrodes, apertures, deflecting electrodes, and objective lens based on the optical conditions and processing and scanning conditions outputted from the data output portion. |
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claims | 1. An x-ray imaging system comprising:a detector positioned to receive x-rays;an x-ray tube coupled to a mount structure and configured to generate x-rays toward the detector, the x-ray tube comprising:a target;a cathode cup;an emitter attached to the cathode cup and configured to emit a beam of electrons toward the target, the emitter having a length and a width; anda one-dimensional grid positioned between the emitter and the target and attached to the cathode cup at one or more attachment points, the one-dimensional grid comprising:a plurality of rungs that each extend in a direction of the width of the emitter;a pair of mounting beams, wherein each of the plurality of rungs comprises at least one end flexibly, slidably or springably attached to a respective one of the mounting beams such that the plurality of rungs are configured to expand and contract relative to the one or more attachment points without substantial distortion with respect to the emitter. 2. The x-ray imaging system of claim 1 wherein each of the plurality of rungs comprises a first end fixedly attached to a first mounting beam and a second end springably attached to a second mounting beam. 3. The x-ray imaging system of claim 1, further comprising:connectors coupled between neighboring pair of rungs; andat least two extension members coupled to the plurality of rungs and configured to attach the plurality of rungs to the cathode cup at respective attachment points, wherein the attachment points are positioned on alternating ends of the rungs, such that the plurality of rungs and their respective connectors form a zig-zag pattern. 4. The x-ray imaging system of claim 1wherein the one-dimensional grid further comprises a plurality of rings forming a coil, each ring forming a rung of the plurality of rungs and configured to encircle the emitter, the coil comprising a pair of legs coupled to the plurality of rings, each leg attached to a respective attachment point. 5. The imaging system of claim 1 wherein the one-dimensional grid further comprises a first mounting beam and a second mounting beam; andwherein a first end of each of the plurality of rungs is flexibly attached to one of the first and second mounting beams and a second end of each of the plurality of rungs is fixedly attached to the other of the first and second mounting beams to allow flexure of the mounting beams along a width direction of the emitter. 6. A method of fabricating a cathode assembly, the method comprising:attaching a filament to a cathode cup;forming a one-dimensional grid having crosspieces that extend generally along a width direction of the filament, wherein forming the one-dimensional grid comprises one of:forming a wire into a zig-zag pattern to form each of the crosspieces, wherein the wire comprises two ends, each end of the wire attached to a respective attachment point; orforming a plurality of coil rings, each coil ring of the plurality of coil rings forming a respective crosspiece of the plurality of crosspieces; orproviding a first support beam and a second support beam, fixedly attaching first ends of the crosspieces to the first support beam, and slideably capturing second ends of the crosspieces in slots in the second beam;positioning the grid proximately to the filament such that electrons that emit from the filament pass between the crosspieces of the one-dimensional grid when accelerated toward an anode; andattaching the grid to the cathode cup at attachment points such that the crosspieces expand, when heated, relative to the attachment points without distorting with respect to neighboring crosspieces. 7. A method of fabricating a cathode assembly, the method comprising:attaching a filament to a cathode cup;forming a one-dimensional grid having crosspieces that extend generally along a width direction of the filament, wherein forming the one-dimensional grid comprises:providing a first support beam and a second support beam; and one offixedly attaching first ends of the crosspieces to the first support beam and springably attaching second ends of the crosspieces to the second beam; orflexibly attaching first ends of the crosspieces to the first support beam and fixedly attaching second ends of the crosspieces to the second beam;positioning the grid proximately to the filament such that electrons that emit from the filament pass between the crosspieces of the one-dimensional grid when accelerated toward an anode; andattaching the grid to the cathode cup at attachment points such that the crosspieces expand, when heated, relative to the attachment points without distorting with respect to neighboring crosspieces. 8. An x-ray tube comprising:a target configured to emit electrons from a focal spot;a cup;an emitter attached to the cup and positioned to emit high-energy electrons toward the focal spot; anda uni-dimensional grated mesh positioned proximately to the emitter and between the target and the emitter such that emitted electrons pass between rungs of the mesh, wherein the uni-dimensional grated mesh comprises a coil and wherein the emitter is positioned within the coil;wherein the uni-dimensional grated mesh is fixedly attached to the cup at attachment points such that rungs of the mesh expand and contract, upon heating and cooling, without substantial distortion with respect to the cup. 9. An x-ray tube comprising:a target configured to emit electrons from a focal spot;a cup;an emitter attached to the cup and positioned to emit high-energy electrons toward the focal spot; anda uni-dimensional grated mesh positioned proximately to the emitter and between the target and the emitter such that emitted electrons pass between rungs of the mesh, the uni-dimensional grated mesh being fixedly attached to the cup at attachment points such that rungs of the mesh expand and contract, upon heating and cooling, without substantial distortion with respect to the cup; anda pair of mounting beams, wherein each of the rungs comprises at least one end flexibly, slidably or springably attached to a respective one of the mounting beams such that the rungs are allowed to expand and contract relative to the one or more attachment points without substantial distortion with respect to the emitter. |
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045253234 | description | DESCRIPTION OF THE INVENTION The present invention is an ion-beam fusion target, described in greater detail hereinafter, but basically consisting of a spherical shell of frozen DT surrounded by a low-density, low-Z pusher seeded with high-Z material, and a high-density, high-Z tamper (see FIG. 1). As pointed out above, this target satisfies many of the requirements or criteria for inertial confinement type commercial power production applications, although it can be also effectively utilized in the other various applications described above. Also, it is again noted that the principal feature of this target is the low-density, low-Z pusher instead of the prior known high-density, high-Z pusher. In addition, this target is composed of inexpensive material and can be readily fabricated by existing technology. The high-density tamper, such as Pb in the illustrated embodiment, serves as a confinement shell to increase the efficiency of the implosion. The pusher material, such as TaCOH, shown in the illustrated embodiment, is CH.sub.2 that has been seeded with a high-Z material, such as tantalum oxide, and is thus a a low-density, low-Z material. The tantalum in the illustrated embodiment constitutes only about 1 atomic percent (at %) of the pusher, but this is sufficient to inhibit energy transport into the fuel, preventing preheat. Because of its low-density, the pusher can be relatively thick to decrease the fluid-instability problem and yet contain little mass. In addition, the fluid instabilities causing pusher-fuel mixing during the final stages of the implosion are ameliorated as a result of the small density difference between fuel and pusher. Calculations typically give Atwood numbers much less than 1 across the pusher-fuel interface during the final stages of the implosion. In some cases, the fuel actually becomes denser than the pusher, resulting in a stable condition. Even where some mixing does occur low-Z materials cause less burn degradation than high-Z materials. In calculations involving high-Z pushers, utilized in previously known targets which are composed of a shell of fuel surrounded by a high-Z pusher material, such as gold (Au), the density times radial thickness (.rho.r) of the pusher in targets comparable to the embodiment illustrated in FIG. 1 is roughly 10 g/cm.sup.2 during thermonuclear burn. In the target of this invention, the bulk of the high-Z material remains uncompressed at a large radius. The total .rho.r of the high-Z material in both the pusher and tamper is less than 1 g/cm.sup.2. Thus, the low-Z pusher target produces less than 10% as much high-Z radioactive debris as a target with a high-Z pusher. Low-Z materials stop ions more effectively than do high-Z materials. Thus the ion-beam energy is preferentially deposited in the pusher region where it is most effective (see FIG. 2) the curve in FIG. 2 was calculated for 6.5 MeV protons at temperatures and densities occurring 18 ns into the implosion, a typical temperature in the deposition media being in excess of 200 eV. To achieve high gain, fuel must be efficiently compressed and only the central portion ignited. A radially propagating burn ignites the remainder of the fuel. These two conditions are achieved by using the pulse shape shown in FIG. 3. FIG. 4 is a plot of pressure as a function of density at the innermost portion of the fuel and at radial distance containing roughly one-half of the fuel. The zero temperature Fermi-degenerate pressure is shown for comparison. It is evident that most of the fuel has been compressed relatively efficiently while the central part of the fuel has been placed on a high adiabat and driven to ignition. Plots of radius as a function of time are shown in FIG. 5. The maximum velocity of the pusher-fuel interface is 33 cm/.mu. sec. Referring now to FIG. 1, the illustrated embodiment of the target comprises a central hollow shell of fuel 10 defining therein a void 11, a pusher shell 12 surrounding fuel shell 10, and a tamper shell 13 surrounding pusher shell 12. As shown in FIG. 1, the fuel 10 is composed of deuterium-tritium (DT) having a density of 0.21 gm/cm.sup.3 mass of 1.00 mg, an inner radius of 0.19004 cm, an outer radius of 0.20000 cm, forming a wall thickness of 0.00996 cm; pusher 12 is composed of TaCOH having a density of 1.26 gm/cm.sup.3, mass of 16.8 mg, inner radius of 0.20000 cm, outer radius of 0.22360 cm, forming a wall thickness of 0.02360 cm; with tamper 13 composed of lead (Pb) having a density of 11.3 gm/cm.sup.3, mass of 72.1 mg., inner radius of 0.22360 cm, outer radius of 0.23333 cm, and wall thickness of 0.009973 cm. While specific parameters and materials of the FIG. 1 embodiment of the target have been described and/or illustrated, the fuel 10 can also be composed of any other thermonuclear fuel such as DD, LiD, etc., but such fuels impose more difficult requirements on ion beam power and energy. The quantity of fuel as well as other parameters such as tamper thickness can range over large limits depending on such things as ion beam voltage, power and energy and desired target yield. The tamper can be composed of any dense material (density of 0.5 to 25 gm/cm.sup.3), for example, 0.5 gm/cm.sup.3 is dense compared to DT. In addition to the above-detailed embodiment the tamper could be composed of materials selected from the group consisting of Pb, Fe, Cu, W, Ag, Ta, Au and any other high Z material. The pusher and intermediate layer can be any low Z material. Seeding material can be nearly anything also. In addition to the above-detailed embodiment the pusher could be composed of materials selected from the group consisting of Ta, COH, Li, Be, B, C or any other low Z material seeded with Ta, W, Pb, Au, Fe, Cu or any high-Z material. The densities of these layers range from 0.07 gm/cm.sup.3 to 10 gm/cm.sup.3. Z of the pusher ranges from 1 to 30 and of the seeding material from 2 to 92, or greater if one wants to use man-made elements. One can also use mixtures of elements. The basic idea is that anything works as long as the tamper is denser than the pusher. The overall gain of the FIG. 1 target is 88. The energy output is 113 MJ; the energy input is 1.28 MJ at a peak power of 2.4.times.10.sup.14 Watts. It is understood that the illustrated target can be readily scaled to different sizes and ion-beam voltages. By way of comparison, the performance of the FIG. 1 target has been compared with a high-Z pusher target composed of a hollow shell of DT having a density of 0.21 gm/cm.sup.3, inner radius of 0.190 cm, and outer radius of 0.200 cm; and a high-Z pusher composed of gold (Au) having a density of 19.3 gm/cm.sup.3, an inner radius of 0.200 cm, and an outer radius of 0.221-0.223 cm; imploded by an unshaped ion beam pulse of 600 TW. A detailed comparison of the FIG. 1 target and the high-Z pusher target is given in Table I. TABLE I ______________________________________ Input Energy Output Energy Peak Input Target Target (MJ) (MJ) Power (TW) Gain ______________________________________ Low-Z 1.28 113 240 88 Pusher High-Z 7.2 247 600 34 Pusher ______________________________________ Two dimensional LASNEX code calculations show the stability of the low-Z pusher to be superior to that of the high Z target. The accuracy of the LASNEX code has been verified by actual implosion experiments (see above-referenced papers UCRL-77056, UCRL-77094, and UCRL-77725), thus unquestionably establishing the code as an effective tool for target design, and thus each target configuration or modification thereof need not be actually imploded to establish the energy produced thereby or energy required to implode same. It has thus been shown that the target of the present invention satisfies many of the requirements for commercial power production by ion-beam fusion, by positioning a layer or shell of low-density, low-Z pusher material seeded with high-Z material between a hollow shell of fuel and a high density tamper. This low-density, low-Z pusher target produces results substantially greater than the targets using a high-Z pusher. While FIG. 1 illustrates a specific embodiment of the invention, it is not intended to limit the invention to the specific materials and parameters illustrated in that embodiment, since as pointed out above, other materials and parameters can be used. Also, if desired for certain applications, the abovedescribed target can be modified by placing a low-density absorber layer between the pusher and the tamper. It has thus been shown that the target of the present invention substantially advances the state of the ion-beam targets for the inertial fusion applications exemplified above. While a particular embodiment of the invention has been described, modifications and changes will become apparent to those skilled in the art, and it is intended to cover in the appended claims all such modifications and changes as come within the spirit and scope of this invention. |
047012803 | summary | The invention provides for a procedure of permanently storing radioactive material in a rock chamber, and particularly for the permanent storage of used nuclear fuel from nuclear reactors and such radioactive waste as is formed by the production of used nuclear fuel (NAGIVA permanent storage). Used nuclear fuel contains uranium, plutonium and fission-products, of which the uranium and the plutonium can be reprocessed and reused as fuel. However, it is not possible today to regain all the uranium and the plutonium, and during reprocessing, waste is formed, which contains, a large number of fission products, small amounts of uranium and plutonium and other transuranic elements. Most of the waste products are extremely radioactive. Since strong radioactive radiation is dangerous to living organisms, it is necessary for the highly active waste to be stored away from such organisms for an extremely long time. It has been suggested that the highly acitve waste be permanently stored under ground, at great depth under primary rock. Such a method of storage would bring about an effective protection from radioactive radiation. However, primary rock normally contains cracks and cavities and often also aquifers. The rock can also be subjected to deformations, for example from earthquakes. With this method of storage, therefore, there is a risk that such deformations of the bedrock can cause the waste containers stored in the rock to break open. Furthermore, there is the risk that the water in the subterranean streams will come into contact with the radioactive waste, which will then be able to spread without control. The radioactive decay also produces heat, causing convection currents in the subterranean streams. In order to reduce the above-mentioned risks, a further method has been suggested in which the radioactive material is stored in a hollow body of solid material, which is placed in a hollow space inside a rock chamber. Such chamber has larger dimensions than the hollow body and the space between the outer casing of the body and the open space is filled with a plastic-deformable material. This space is in turn surrounded by a further, outer spacing surrounding the first space on all sides; this space must likewise be filled with a plastic-deformable material. The plastic material must have low permeability for water and must not split when deformed. The encapsulated material is stored freely within the hollow body for insertion and removal of the material and there are inspection-holes in the body for monitoring. However, this construction does not provides an effective permanent method of storage for all time, and, with its provision of several inspection shafts and the storage of material capsules lying freely inside the body, it can hardly be considered to fulfil the requirements for permanent storage, i.e. a period of thousands of years. This method would require considerable expense, both for construction and inspection. The present invention, the characteristics of which are described in the patent claims, provides a method for permanent storage of radioactive material, which is both safe and requires no maintenance but which can, if so desired, be continuously inspected and allows any faults to be rectified. |
abstract | A boiling water nuclear reactor comprises: a reactor containment vessel including a dry well and a wet well; a vent pipe connecting the dry well and the pressure suppression pool; a gravity-driven water injection pool to hold boric acid aqueous solution; an emergency core water-injection piping system for causing the boric acid aqueous solution in the gravity-driven water injection pool to fall so as to be injected into the reactor pressure vessel in case of reactor accident; a static containment vessel cooling system pool; a static containment vessel cooling system heat exchanger; a dry well connection pipe connecting an upper part of the static containment vessel cooling system heat exchanger and the dry well; and a gas vent pipe for discharging noncondensible gas in the static containment vessel cooling system heat exchanger into the inside of the pressure suppression pool. |
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043702953 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is a block diagram of prior art toroidal-type fusion-fission reactor designs. The figure illustrates a cross-sectional representative view of the toroid which is symmetric about a major axis 24. Typically, a toroidal fusion region 4 is provided for housing the fusion fuel which may be, for example, a mixture of deuterium and tritium. Surrounding the fusion region 4 is a fission blanket 6 which contains fissionable material such as natural uranium (U.sup.238) or a uranium alloy. A moderator/reflector region 8 is also provided around the fission blanket 6 to slow down fast neutrons produced in both the fusion and fission regions and to reflect these neutrons as low energy neutrons back into the fission blanket 6. The resulting thermal neutrons within the fission blanket do not induce fission reactions in U.sup.238 but rather are captured and lead to the production of fissile material, namely, Pu.sup.239. After the moderator/reflector region 8 there is generally provided a T breeding section 10 which comprises lithium utilized to breed tritium via thermal and/or fast neutron capture. The tritium may then be utilized to replace tritium consumed by the d,t fusion reaction. Surrounding the T breeding section 10 is a shielding area 12. Toroidal field (TF) coils 14 surround and are protected by the shielding area 12 and are utilized to generate a toroidal field within the fusion region 4. Ohmic heating (OH) coils 16 are also illustrated adjacent the toroidal field coils 14 and are utilized to ohmically heat the plasma fusion region 4. FIG. 2A is a schematic block diagram of a toroidal fusion device similar to FIG. 1 but illustrating the novel structural arrangement in accordance with the principles of the invention. Accordingly, a fusion-fission reactor 20 is illustrated comprising a fusion region 22 which is typically in the form of a toroid having a main axis 24. A plurality of toroidal field coils 24' are provided which have incorporated within a region thereof the fissile-fertile material 26. The fissile-fertile material 26, may, of example, comprise natural uranium, a uranium-molybdenum or uranium-zirconium alloy or the like. The fissile-fertile material is fissionable with respect to high energy neutrons produced from the fusion reaction and is fertile with respect to low energy neutrons thus producing additional fissile material. The particular structure of the toroidal field coils 24' and the fissile-fertile material 26 is described more fully below and may typically comprise an integral structure positioned substantially adjacent the fusion region 22 with only a vacuum vessel (and associated cooling panels) therebetween to house the plasma and cool the vacuum vessel first wall. Surrounding the toroidal field coils 24' and fissile-fertile material 26 is a containment means 28 for housing the toroidal field coils 24'. The containment means 28 may be fabricated from stainless steel, copper or other metals and is provided with an insulative coating to prevent shorting of the TF coils. Also provided within the containment means 28 are the ohmic heating (OH) coils 36. Provisions are made for connecting the OH and TF coils to appropriate power supplies as is well known in the art. A moderator/reflector region 30, fabricated from graphite, for example, is also provided exterior to the containment means, and a T breeding section (lithium) 32 is positioned exterior of the moderator/reflector region 30. A shielding area 34 is also provided around the T breeding section 32. FIG. 2B is another embodiment of the invention similar to FIG. 2A but shows the containment means 28b surrounding the TF coils 24 and fusion region 22 with the OH coils 36 outside thereof. OH coils 36 may also be enclosed in a separate containment means, not shown. FIG. 3 illustrates yet another embodiment of the invention which is similar to that of FIG. 2 with the exception that the OH coils 36' are now positioned within the region of the toroidal field coils generally disposed in-between the fusion area 22' and the TF coils 34'. Elements similar to those in FIG. 2 are labeled with primes. The embodiment of FIG. 3 essentially frees the interior of the toroidal area near the main axis 24' of the traditional bulky OH coil transformer and provides for the efficient OH heating of the fusion region 22' by positioning the OH coils substantially adjacent the fusion region 22'. Additionally, the removal of the OH coils from the interior region of the toroid permits a more efficient design of the TF coils 24' by allowing for larger TF coil cross-sectional areas within this vicinity of the main axis 24 with substantial reduction in current densities. A resulting increase in TF magnetic strength is achieved which, coupled to a more efficient operation of the OH coils, enhances efficiency and stability of the operation of the fusion region 22'. Positioning the OH coils 36' within the region of the TF coils 24' is applicable to pure fusion toroidal-type reactors as well as the fusion-fission type. In operation of the fusion-fission power generating device there is typically provided a plurality of fusion-fission reactors as per FIGS. 2 or 3 interconnected in a power generating plant. Such a system is schematically depicted in FIG. 4. FIG. 4 is illustrated for the fusion-fission unit 20 of FIG. 2 although it is readily understood that a similar system could be operated with respect to the corresponding elements of FIG. 3. The containment means 28 together with everything contained therein, in particular, the OH coils 36, the toroidal field coils 24', the fissile-fertile material 26 and the fusion region 22, form a unit which may be identified as a fusion-fission power core (FFPC) 40. The FFPC core 40 is separable from the remaining elements of the fusion-fission reactor, namely, the moderator/reflector 30, T breeding section 32 and shielding area 34. A unique advantage in accordance with the principles of the invention is that each of the fusion-fission reactors 20 is modular in the sense that the FFPC 40 may effectively be separated and removed from the remaining reactor elements for replacement by a replacement or substitute FFPC after the fissile-fertile material has been depleted to the extent that fission reactions no longer contribute in an optimum way to the overall energy production of the machine. Depending upon the operating parameters of the reactor 20, the fusion-fission power core 40 may be replaced on the order of yearly intervals. FIG. 4 generally depicts the interconnection of the fusion-fission reactor to power supply means 50 and heat exchange, pumps and turbine apparatus 54. Power supply 50 is utilized to provide power to the OH and TF coils as well as power to auxiliary heating and equilibrium field coils (not shown) as are present in conventional Tokamak designs. The heat exchange, pumps and turbine apparatus 54 are generally interconnected to the fusion-fission power core 40 to extract thermal energy therefrom and simultaneously to cool the fusion-fission core during operation thereof. The overall interconnection of the fusion-fission power cores in a modular array to form a power generating plant may be similar to the modular fusion apparatus more fully illustrated in copending application entitled "Modular Fusion Power Apparatus Using Disposable Core", Ser. No. 841,903, incorporated herein by reference. A more detailed illustration of the invention is shown in FIGS. 5-7. These figures depict the embodiment of FIG. 3 wherein the OH coils 36' are positioned interior of the toroidal field coils 24'. However, it is readily understood that the OH coils may be positioned as illustrated in FIG. 2 exterior of the TF coils as shown, for example, in the copending application referenced hereinabove. The structure of the toroidal field coils and the fissile-fertile material is substantially the same with the exceptions that the cut-out region for the OH coil within the TF coil structure is no longer present, and that the radial dimension of the TF coils toward the main axis of the toroid is reduced in order to allow space for the OH transformer in the region of the toroid main axis. FIG. 5 is a top plan view of the portion of the fusion-fission reactor 20' wherein the main axis and center of the toroid is to the left in the drawing. The reactor comprises a plasma containment means 100 utilized to contain the fusion plasma gas, typically a mixture of deuterium and tritium. It is understood, however, that in addition to the d,t reaction additional fusion reactions may be utilized such as D,D or D,He.sup.3 etc. The plasma containment means 100 is symmetric about the main axis 102. In both FIGS. 5 and 6, however, the main axis 102 is shown closer to the plasma containment means 100 than dictated by the scale factors of the drawing in order to illustrate in clear detail the novel toroidal field structure of the invention. Surrounding the plasma containment means 100 is a toroidal field generating means generally indicated at 104 which comprises a plurality of TF sectors 106 each of which comprises the toroidal field coils 108. The main axis of the toroid defined by TF sectors 106 will typically coincide with the main axis of the containment means 100. FIG. 5 illustrates the OH coils 110, containment means 112, moderator/reflector 114 and T breeding section 116. The modulator/reflector 114 is shown segmented to indicate that it would typically be of a size larger than illustrated and would generally be on the order of the TF coil radius. Particular cooling and cladding of the modulator/reflector as well as the structure of the T breeding section 116 are well known in the art and are not shown for the sake of clarity. In accordance with the novel aspects of the invention, the toroidal field sector 106 is seen to comprise a region of fissile-fertile material 120 such as, for example, natural uranium, or any of the uranium alloys such as uranium molybdenum, zirconium, etc. The fissile-fertile region within each of the TF sectors 106 is shown in the form of two separate regions namely, regions 120a and 120b, which effectively sandwich a portion of the TF coils 108 therebetween. Other arrangements are of course possible such as positioning the region 120 entirely on one side of the TF sectors 106. In reference to FIG. 6 it may be seen that the fissile-fertile material 120 does not extend completely around the TF sector 106 but rather is disposed primarily in the region away from the main axis 102. FIG. 6 illustrates that the fissile-fertile material region 120 is disposed around the TF sector 106 through an angle .theta. which may typically be on the order of 240.degree.. It is additionally seen that the regions 120a and 120b optimumally extend to a region as close to the fusion region as possible to take maximum advantage of the neutron flux. Typically, the regions 120a and 120b extend from substantially the inner circumferential contour of the toroidal field coils 108 to substantially the outer circumferential contour thereof. These contours are shown by indicia C1 and C2 respectively. The regions of fissile-fertile material 120a and 120b are encased in a cladding 122 which may typically be of stainless steel wherein there is provided a plurality of cooling channels (or grooves) 124. Cooling channels 124 are likewise provided throughout the TF coils 108, a few of which channels are illustrated in FIGS. 6 and 7. Because of the close proximity of the TF coils 108 to the plasma containment means 100, the entire region of the TF coils typically contains either radial cooling channels, as illustrated, or circumferential cooling channels if desired. These channels may be spaced on the order of a centimeter apart and may typically occur both within the TF coils 108 and on the cladding 122. As best illustrated in FIG. 6, these cooling channels may also be in fluid communication with a cut-out region 130 of the TF coils 108, to thereby cool the OH coils 128 which are illustrated as being inserted within cut-out region 130. As a representative example of the dimensions of the apparatus of FIGS. 5-7, the toroid radius to the center of the plasma may be on the order of 50-100 cm, the plasma radius 20-50 cm, moderator/reflector region thickness 10-30 cm and the lithium T breeding section a distance from the center of the plasma on the order of 300 cm. The volume fraction of U.sup.238 to Cu in the TF coils may typically be from 5-50%. For a plasma center taken as the origin, calculations using a cylindrical model geometry have shown a favorable selection of parameters as follows: plasma radius, approximately 23 cm; inner and outer radius of U.sup.238 -Cu region of TF coil, 23 cm and 33 cm respectively; inner and outer radius of moderator/reflector region, approximately 33 cm and 48 cm respectively; and lithium T breeding section extending from 300 cm to approximately 400 cm from the origin. The volume fraction of U.sup.238 to copper is optimumally 20%, and number densities for U.sup.238 and Cu may be taken as 8.7.times.10.sup.21 /cm.sup.3 and 6.3.times.10.sup.22 /cm.sup.3 respectively. Particular details in regard to the structure of cooling channels for blanket regions as well as TF coil structures per se have been the subject of many studies in the prior art and reference is made to the aforementioned Review Meeting and Proceedings for additional details with respect thereto. Additional prior art toroidal coil designs are illustrated in U.S. Pat. Nos. 3,859,615 and 3,303,449 and cooling configurations for blanket regions and the like are well known in the art, such as, for example, Volume I and II of Tokamak Experimental Power Design Conceptual Design, Argon National Laboratory, August 1976, ANL/CTR-76-3. Additionally, it is pointed out that the plasma containment region 100 although illustrated as a simple shell in practice may contain a coolant panel in fluid communication with the radial or circumferential cooling channels 124 within the TF coils 108 and cladding 122. Reference is made to the aforementioned ANL publication as representative of a typical design in relation to the vacuum vessel and coolant panel corresponding to the plasma containment means 100. Although the plasma containment means 100 is illustrated as having a circular cross-sectional area, it is understood that additional designs recognized in the art are also possible, such as, for example, the D configuration well known in Tokamak studies. The corresponding TF generating means 104 is naturally designed consistent with the shape of the plasma containment means 100. Of particular significance with respect to the invention is, however, that the TF coils 10 be positioned substantially adjacent (consistent with first wall loading and cooling reguirements) to the plasma containment means 100 and additionally that the region of fissile-fertile material 120 is also positioned substantially adjacent to the plasma containment means 100. This particular arrangement allows optimization of the fusion-fission reaction and produces an extremely large burn up of the uranium fuel. This close proximity of the TF coil sectors 106 and the fissile-fertile region 120 to the fusion area permits a high fission-to-fusion energy production ratio on the order of 10:1. This high ratio of fission to fusion energy allows operation of the FFPC at gross fusion power levels significantly less than those which would be required for operation of a pure fusion power device--in particular, FFPC lifetimes, as limited by radiation damage by fusion neutrons, may be ten or more times greater for the same net nuclear energy output (fission plus fusion) than for fusion alone. Thus, this invention allows operation over FFPC lifetimes as high as one to two years. Over such long time periods the high fusion neutron flux will result in large burnup utilization of the U.sup.238 in the coils, in proximity to the plasma. Indeed, calculations indicate that as much as 80% of the uranium should be fissioned in the first three centimenters of the fissile-fertile region 120 nearest the plasma containment means 100. Average burn up percentages are on the order of 50%. In operation of the fusion-fission reactor or power generating means, high energy neutrons on the order of 14 Mev are generated by the d,t reaction within the plasma containment means 100. These neutrons cause fission of U.sup.238 within the regions 120. The fission reactions generate fission fragments plus large amounts of energy on the order of 200 Mev/fission. Additionally, the fission reactions caused by fusion neutrons generate up to 4.5 neutrons per fission with neutron energies in the range of 1-5 Mev. These neutrons in turn generate slight additional fission reactions with U.sup.238. Neutrons which escape the TF sectors 106 are slowed down in the moderator/reflector 114 and are refelected back into the fissile-fertile regions 120 wherein these low energy neutrons are captured by U.sup.238 which eventually decays to Pu.sup.239. The Pu.sup.239, in turn, is fissile and thus fissions upon reactions with thermal neutrons (as well as fast neutrons). Pu.sup.239 eventually reaches a saturation level which is sufficient to contribute significantly to the overall energy production of the reactor. The particular composition and thickness of the moderator/reflector is selected to slow down and reflect neutrons into the fissile-fertile material at energies which optimize overall energy production of the reactor. The terminology "low energy neutrons" as utilized in the appended claims thus is intended to cover such neutron energies. For example, epithermal neutrons are expected to have large cross sections for n capture, and known resonance peaks in U.sup.238 extend generally in the range of 6-200 ev. Thermal neutrons will also contribute to Pu.sup.239 production. An important aspect of the design of the fusion-fission reactor in accordance with the principles of the invention is in the positioning of the TF coils substantially adjacent the toroidal fusion region. Thus, while there is typically a vacuum chamber containing the plasma of the fusion region and while there may generally be a cooling region provided to cool the chamber first wall, the TF coils are positioned adjacent any such cooling region such that they are substantially adjacent to the fusion region itself. The positioning of the TF coils substantially adjacent the fusion region and the provision for positioning the fissile-fertile material in the region of the TF coils is a sharp contrast to prior art designs. It is clear that the terminology of placing the fissile-fertile material "within the region of the TF coils" does not require that the material coincide in a spatial sense with the electrically conductive coil material (copper, for example). The fissile-fertile material typically will be a distinct region positioned within the region defined between the inner and outer circumferential contour of the TF coils. With respect to the OH coils, it is clear that these coils likewise are positioned substantially adjacent the fusion region but are located on the side of the toroidal fusion region nearest the main axis thereof. Although the invention has been described in terms of selected preferred embodiments, the invention should not be deemed limited thereto since other embodiments and modifications will readily occur to one skilled in the art. It is therefore to be understood that the appended claims are intended to cover all such modifications as fall within the true spirit and scope of the invention. |
044316038 | claims | 1. In a nuclear reactor having a nuclear fuel assembly therein with coolant flowing therethrough, said coolant flow passing through an orifice, the improvement comprising neutron fluence responsive, self-actuated flow control valve for the orificing of coolant flow through said nuclear fuel assembly which valve comprises: (a) a shaft; (b) plug means for controlling the coolant flow through said orifice, said plug means connected to said shaft for actuation thereby; (c) spring means for biasing said plug means to a predetermined position; (d) a gas-tight bellows connected to said shaft such that expansion of said bellows moves said shaft thereby actuating said plug means; (e) a mass containing a material chosen from the group boron, lithium, or beryllium, said mass located within said bellows such that gas released by a nuclear reaction of said material with neutrons accumulates in said bellows, said mass chosen in quantity so as to match the actuation of said valve with a chosen neutron fluence. |
abstract | A method for analysing at least one fuel rod comprising a stack of nuclear fuel, a rod comprising packed zones completely filled with fuel and intermediate zones partially full of fuel, comprises: acquiring a count profile associated with a non-migrating isotope, a profile being made up of spectrometry measurements taken along the rod for this isotope; determining a set of at least one indicator K_i that makes it possible to quantify the reduction in material at an intermediate zone of index i, the said indicator being deduced from the count profile; detecting the change in geometry by comparing the set of at least one indicator K_i against a set of at least one reference value RK indicative of the initial geometry of the nuclear fuel stack. |
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abstract | Steering system for a droplet generator in a EUV system. The steering system permits controlled positioning of a droplet release point of the droplet generator. A movable member holding the droplet generator is coupled to stationary elements of the EUV system through a coupling system having a first subsystem that constrains lateral translation of the movable member, and a second subsystem that controls a relative inclination of the movable member. The first and second subsystems preferably include one or a combination of flexures that permit highly precise and repeatable positioning. |
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summary | ||
054992761 | abstract | Neptunium of minor actinide nuclides separated from spent fuel is added to fuel of reactor cores (inner reactor cores and/or outer reactor cores) of a fast reactor and americium of the separated minor actinide nuclides and rare earth elements are added to either or both of radial and axial blankets of the fast reactor for burning. Thus, the minor actinide nuclides with long half-lives can be burnt with the fast reactor core with the minimized effects of the rare earth elements. For a burner reactor, americium and rare earth elements may be added to shields for burning. Curium may be added together with americium and rare earth elements. Neptunium is added in amount of 2% to 5% by weight based on the weight of the fuel and the rare earth elements are added in an amount of 50% by weight or less based on the weight of the fuel. A Purex process is used to separate neptunium and a Truex process is used to separate americium and curium. |
summary | ||
summary | ||
description | The present application is a continuation of U.S. patent application Ser. No. 14/398,946 filed Nov. 4, 2014, which is a national stage application under 35 U.S.C. § 371 to international application No. PCT/US2013/039743 filed May 6, 2013, which claims priority to U.S. Provisional Patent Application No. 61/642,614, filed May 4, 2012; the disclosures of which are incorporated herein by reference in their entireties. The field of the present invention relates to nuclear steam supply systems, and more particularly to a steam supply system for small modular reactors. Pressurized water reactors (PWRs) for nuclear power generation facilities utilize both pumped and natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant to produce steam which may be working fluid for a Rankine power generation cycle. The existing natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service, but also subjects the equipment to corrosive conditions and results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. In addition, locating the heat exchange equipment within the reactor pressure vessel creates problems with respect to radiation levels encountered for crews to repair the heat exchange equipment in proximity to the radioactively hot components of the reactor vessel. The general view has also been that the heat exchangers should be located in the reactor vessel to achieve natural circulation in those systems which may utilize this type of flow circulation. The reduction of vulnerabilities within nuclear power generation facilities is always an ongoing issue. For example, large pipes are seen as creating the potential for a “large break” Loss of Coolant Accident (LOCA) event, and thus it is desirable to remove large pipes where possible. A nuclear reactor vessel includes a shell and a head affixed to the upper end of the shell. The shell has an internal cavity with a central axis and an upper flange portion, wherein the internal cavity is configured to receive a reactor core. The head has a head flange portion, with the upper annular flange portion is coupled to the head annular flange portion, and the flanges are configured to minimize outward extension from the cavity while still providing desired leak protection at the interface between the shell and the head. In a first separate aspect of the present invention, the upper flange portion of the shell is annular and extends into the internal cavity, and the head flange portion of the head is also annular and extends outward from the internal cavity. In a second separate aspect of the present invention, a reactor core including nuclear fuel is disposed within the internal cavity of the nuclear reactor vessel, and a steam generating vessel including at least one heat exchanger section is fluidicly coupled to the reactor vessel. The upper flange portion of the shell extends into the internal cavity, and the head flange portion of the head extends outward from the internal cavity. In a third separate aspect of the present invention, a reactor core including nuclear fuel is disposed within the internal cavity of the nuclear reactor vessel, and a steam generating vessel including at least one heat exchanger section is fluidicly coupled to the reactor vessel. The upper flange portion of the shell extends into the internal cavity, and the head flange portion of the head extends outward from the internal cavity. An inner surface of the first head portion is disposed closer to the central axis than an inner surface of the first shell portion along respective parallel radial lines extending from the central axis. In a fourth separate aspect of the present invention, a method for generating steam utilizes the nuclear reactor vessel. The reactor vessel is capped with a head, and a reactor core is disposed within the reactor vessel. The upper flange portion extends into the internal cavity, and the head flange portion extends outward from the internal cavity. A liquid primary coolant is heated in the nuclear reactor core, and the heated primary coolant is discharged from a top portion of the reactor vessel into a steam generating vessel. The primary coolant is flowed through the reactor vessel and steam generating vessel in a closed circulation loop. In a fifth separate aspect of the present invention, one or more of the preceding separate aspects may be employed in combination. Advantages of the improvements will be apparent from the drawings and the description of the embodiments below. The description of illustrative embodiments according to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments of the invention disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,” “above,” “below,” “up,” “down,” “left,” “right,” “top” and “bottom” as well as derivatives thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation unless explicitly indicated as such. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Moreover, the features and benefits of the invention are illustrated by reference to the preferred embodiments. Accordingly, the invention expressly should not be limited to such preferred embodiments illustrating some possible non-limiting combinations of features that may exist alone or in other combinations of features; the scope of the invention being defined by the claims appended hereto. Referring to FIGS. 1-6, a steam supply system for a nuclear pressurized water reactor (PWR) according to the present disclosure is shown. From the thermal-hydraulic standpoint, the system includes a steam generator assembly 100 generally including a reactor vessel 200 and a steam generating vessel 300 fluidly coupled to the reactor vessel. The steam generating vessel and reactor vessel are vertically elongated and separate components which hydraulically are closely coupled, but discrete vessels in themselves that are thermally isolated except for the exchange of primary loop coolant (i.e. reactor coolant) flowing between the vessels. As further described herein, the steam generating vessel 300 in one embodiment includes a preheater 320, main steam generator 330, and a superheater 350 which converts a fluid such as water flowing in a secondary coolant loop from a liquid entering the steam generating vessel 300 at an inlet 301 to superheated steam leaving the steam generating vessel at an outlet 302. The secondary coolant loop water may be a Rankine cycle fluid used to drive a turbine-generator set for producing electric power in some embodiments. The steam generating vessel 300 further includes a pressurizer 380 which maintains a predetermined pressure of the primary coolant fluid. The pressurizer is a pressure vessel mounted atop the steam generating vessel 300 and engineered to maintain a liquid/gas interface (i.e. primary coolant water/inert gas) that operates to enable control of the primary coolant pressure in the steam generator. In one embodiment, as shown, the pressurizer 380 may be mounted directly on top of the steam generating vessel 300 and forms an integral unitary structural part of the vessel to hydraulically close the vessel at the top end. The assemblage of the foregoing three heat exchangers and the pressurizer may be referred to as a “stack.” Referring to FIG. 1, the reactor vessel 200 and the steam generating vessel 300 are housed in a steam generator containment vessel 110. The containment vessel 110 may be formed of a suitable shop-fabricated steel comprised of a top 111, a bottom 112, and a cylindrical sidewall 113 extending therebetween. In some embodiments, portions of the containment vessel which are located above ground level may be made of ductile ribbed steel to help withstand aircraft impact. A missile shield 117 which is spaced above the top 111 of the containment vessel 110 may be provided as part of the containment vessel or a separate containment enclosure structure (not shown) which encloses the containment vessel 110. A horizontal partition wall 114 divides the containment vessel into an upper portion 114a and a lower portion 114b. The partition wall 114 defines a floor in the containment vessel. In one embodiment, a majority of the reactor vessel 200 may be disposed in the lower portion 114b and the steam generating vessel 300 may be disposed in the upper portion 114a as shown. In various embodiments, the containment vessel 110 may be mounted above ground, partially below ground, or completely below ground. In certain embodiments, the containment vessel 110 may be positioned so that at least part or all of the lower portion 114b that contains the nuclear fuel reactor core (e.g., a fuel cartridge 230) is located below ground level. In one embodiment, the entire reactor vessel 200 and a portion of the steam generating vessel 300 are located entirely below ground level for maximum security. The cylindrical shell or sidewall 113 of the containment vessel 110 may be horizontally split into an upper section and a lower section, which are joined together by a circumferential welded or bolted flanged joint 119 as shown in FIG. 1 to provide a demarcation for portions of the containment vessel which are located above and below ground level. In other embodiments, the upper and lower sections may be welded together without use of a flange. In one embodiment, for example without limitation, the containment vessel 110 may have a representative height of approximately 200 feet or more for a 160 MW (megawatt) modular nuclear electric generation facility. A non-limiting representative diameter for this power generation facility is about 45 feet. Any suitable height and diameter for the containment vessel may be provided depending on system component configuration and dimensions. The containment vessel 110 further includes a wet reactor well 115 defined in one embodiment by a cylindrical circumscribing walled enclosure 116 which is flooded with water to provide enhanced radiation shielding and a back-up reserve of readily accessible coolant for the reactor core. In one embodiment, the walled enclosure 116 may be formed of stainless steel cylindrical walls which extend circumferentially around the reactor vessel 200 as shown. Other suitable materials may be used to construct the enclosure 116. The wet reactor well 115 is disposed in the lower portion 114b of the containment vessel 110. The lower portion 114b may further include a flooded (i.e. water) used fuel pool 118 adjacent to the enclosure 116. In one embodiment, as shown in FIG. 1, both the used fuel pool 118 and the walled enclosure 116 are disposed below the horizontal partition wall 114 as shown in FIG. 1. In one embodiment, as shown in FIG. 1, the walled enclosure 116 may extend above the partition wall 114 and the inlet/outlet nozzle connection between the reactor and steam generating vessels may be made by a penetration through the walled enclosure. As further shown in FIG. 1, both the reactor vessel 200 and the steam generating vessel 300 preferably may be vertically oriented as shown to reduce the footprint and diameter of the containment vessel 110. The containment vessel 110 has a diameter large enough to house both the reactor vessel, steam generating vessel, and any other appurtenances. The containment vessel 110 preferably has a height large enough to completely house the reactor vessel and steam generating vessel to provide a fully contained steam generator with exception of the water and steam inlet and outlet penetrations for second coolant loop fluid flow associated with the Rankine cycle for driving the turbine-generator set for producing electric power. FIG. 2 shows the flow or circulation of primary coolant (e.g. water) in the primary coolant loop. In one embodiment, the primary coolant flow is gravity-driven relying on the change in temperature and corresponding density of the coolant as it is heated in the reactor vessel 200, and then cooled in the steam generating vessel 300 as heat is transferred to the secondary coolant loop of the Rankine cycle which drives the turbine-generator (T-G) set. The pressure head created by the changing different densities of the coolant (i.e. hot—lower density and cold—higher density) induces flow or circulation through the reactor vessel-steam generating vessel system as shown by the directional flow arrows. Advantage, the gravity-driven primary coolant circulation requires no coolant pumps or machinery thereby resulting in cost (capital, operating, and maintenance) savings, reduced system power consumption thereby increasing energy conversion efficiency of the PWR system, in addition to other advantages as described herein. The reactor vessel 200 may be similar to the reactor vessel with gravity-driven circulation system disclosed in commonly-owned U.S. patent application Ser. No. 13/577,163 filed Aug. 3, 2012, the disclosure of which is incorporated herein by reference in its entirety. Referring to FIGS. 3A and 3B, the reactor vessel 200 in one embodiment is an ASME code Section III, Class 1 thick-walled cylindrical pressure vessel includes a cylindrical sidewall shell 201, an integrally welded hemispherical bottom head 203 and, a removable hemispherical top head 202. The shell 201 primarily defines an internal cavity 208 configured for holding the reactor core, reactor shroud, and other appurtenances as described herein. In one embodiment, the upper extremity of the reactor vessel shell 201 is equipped with a tapered hub flange 204 (also known as “welding neck” flange in the art) which is bolted to a similar flange 205 welded to the top head 202. Commonly-owned PCT patent application No. PCT/US2013/0038289, filed Apr. 25, 2013, the disclosure of which is incorporated herein by reference in its entirety, discloses known prior-art for the design and coupling of the top head to the shell using two flanges. Each flange 204, 205 may be annular, so that each extends completely around the shell 201 and the head 202, respectively. Each flange may also be integrally formed as part of the shell 201 and the top head 202. The flange 204 extends into and toward the central axis 209 of the cavity 208, with the flange 204 forming at about the point where the sidewalls of the shell 201 begin to widen. In the case that the flange 204 is annular, it extends radially into the cavity around the entire cavity, and similarly, in the case that the flange 205 is annular, it extends radially outward from the cavity all around. As can be seen in the embodiment depicted, the inner sidewall surfaces 209 of the shell 201, excluding the flange 204, are defined by a first inner radius, measured from the central axis 210 of the cavity 208, and the inner sidewall surfaces 211 of the flange 204 are defined by a second inner radius, with the second inner radius being smaller than the first inner radius. The outward-extending head flange 205, which is formed at about the point where the sidewalls of the top head 202 begin to widen, has an inner surface 212 that may be at about the same distance from the central axis 210 as the inner sidewall surfaces 211 of the shell flange 204, as can be seen by line A. Thus, the inner surfaces 211, 212 of the two flanges 204, 205 have about the same radius from the central axis 210, and the inner surfaces 212 of the flange 205 have a smaller radius than the inner sidewall surfaces 209 of the shell 201. Also, the outer surfaces 213 of the hemispherical wall of the top head 202, at a point just above the flange 205, may be at about the same distance from, or even closer to, the central axis 210 as the inner sidewall surfaces 211 of the shell flange 204, as can be seen by line B. This results in the radius of the outer surfaces 213 having about the same radius from the central axis 210 as the inner sidewall surfaces 211, although the outer surfaces 213 could also have a radius less than that of the inner sidewall surfaces 211. So that the coupled flanges 204, 205 may each still serve as a “welding neck” flange, the outer surfaces 214, 215 of each flange may be at about the same distance from the central axis 210, as can be seen by line C. All distance and measurement comparisons between the shell 201 and the top head 210 are being made along parallel radial lines having the central axis 210 as a center. The top head 202 may be fastened to the shell 201 by coupling the flanges 204, 205 via a set of alloy bolts 216, which are pre-tensioned to establish a high integrity double gasket seal under all operation modes. The bolted connection of the top head 202 provides ready access to the reactor vessel internals such as the reactor core. The centerline, line D, of the bolts 216 may be at a distance greater than the internal surfaces 209 of the shell 201, but at a lesser distance than the outer surfaces 214 of the shell flange 204. In one embodiment as shown in FIG. 3B, the outer surface 214 of the shell flange 204 is substantially flush with the vertical outer surface of the cylindrical shell 201 of the reactor vessel 200 immediately below the upper flange portion such that there is no appreciable projection of the flange beyond the shell. Two concentric self-energizing gaskets 206 are placed in a pair of annular grooves 218, the grooves being formed in both flanges 204, 205, between the bolts 216 and the inner surfaces 211, 212, and compressed between the interfacing surfaces of two flanges 204, 205, when coupled together, to provide leak tightness of the reactor vessel 200 at the connection between the top head 202 and the shell 201. The leak tightness under operating conditions is assured by an axisymmetric heating of the flanged joint that is provided by the fluid flow arrangement of the primary coolant in the system, as further described herein. The top head 202 contains the vertical penetrations 207 for insertion of the control rods and further may serve as a base for mounting the associated control rod drives, both of which are not depicted but well known in the art without further elaboration. With continuing reference to FIG. 3A, the reactor vessel 200 includes a cylindrical reactor shroud 220 which contains the reactor core defined by a fuel cartridge 230. The reactor shroud 220 transversely divides the shell portion of the reactor vessel into two concentrically arranged spaces: (1) an outer annulus 221 defining an annular downcomer 222 for primary coolant entering the reactor vessel which is formed between the outer surface of the reactor shroud and the inner surface of the shell 201; and (2) a passageway 223 defining a riser column 224 for the primary coolant leaving the reactor vessel heated by fission in the reactor core. The reactor shroud 220 is elongated and extends in an axial direction along vertical axis VA1 of the reactor vessel which defines a height and includes an open bottom 225 and a closed top 226. In one embodiment, the top 226 may be closed by a top flow isolation plate 227 which directs primary coolant flowing up the riser column 224 to the steam generating vessel 300, as further described herein. In one embodiment, the bottom 225 of the reactor shroud 220 is vertically spaced apart by a distance from the bottom head 203 of the reactor vessel 200 and defines a bottom flow plenum 228. The bottom flow plenum 228 collects primary coolant from the annular downcomer 222 and directs the coolant flow into the inlet of the riser column 224 formed by the open bottom 225 of the reactor shroud 220 (see, e.g. FIG. 2). Both the fuel cartridge 230 and the reactor shroud 220 are supported by a core support structure (“CSS”), which in one embodiment includes a plurality of lateral support members 250 that span between and are attached to the reactor shroud and the shell 201 of the reactor vessel 200. A suitable number of supports members space both circumferentially and vertically apart are provided as needed to support the combined weight of the fuel cartridge 230 and the reactor shroud 220. In one embodiment, the bottom of the reactor shroud 220 is not attached to the reactor vessel 200 to allow the shroud to grow thermally in a vertical axial direction (i.e. parallel to vertical axis VA1) without undue constraint. The reactor shroud 220 is a double-walled cylinder in one embodiment 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 (i.e. water) heated by the fission in the fuel cartridge 230 (“core”), which is preferably located at the bottom extremity of the shroud in one embodiment as shown in FIGS. 1-3. 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. The fuel cartridge 230 in one embodiment is a unitary autonomous structure containing upright fuel assemblies, and is situated in a region of the reactor vessel 200 that is spaced above the bottom head 203 so that a relatively deep plenum of water lies underneath the fuel cartridge. 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 and adjoining upper portions of the riser column 224. The fuel cartridge 230 is an open cylindrical structure including cylindrically shaped sidewalls 231, an open top 233, and an open bottom 234 to allow the primary coolant to flow upward completely through the cartridge (see directional flow arrows). In one embodiment, the sidewalls 231 may be formed by multiple arcuate segments of reflectors which are joined together by suitable means. The open interior of the fuel cartridge 230 is filled with a support grid 232 for holding the nuclear fuel rods and for insertion of control rods into the core to control the fission reaction as needed. Briefly, in operation, the hot reactor primary coolant exits the reactor vessel 200 through a low flow resistance outlet nozzle 270 to be cooled in the adjacent steam generating vessel 300, as shown in FIGS. 2 and 3. The cooled reactor primary coolant leaves the steam generating vessel 300 and enters the reactor vessel 200 through the inlet nozzle 271. The internal plumbing and arrangement in the reactor vessel directs the cooled reactor coolant down through to the annular downcomer 222. The height of the reactor vessel 200 is preferably selected to support an adequate level of turbulence in the recirculating reactor primary coolant by virtue of the density differences in the hot and cold water columns which is commonly known as the thermo-siphon action (density difference driven flow) actuated by gravity. In one embodiment, the circulation of the reactor primary coolant is driven by over 8 psi pressure generated by the thermo-siphon action, which has been determined to ensure (with adequate margin) a thoroughly turbulent flow and stable hydraulic performance. Referring to FIGS. 1 and 3, the top of the reactor vessel shell 201 is welded to a massive upper support forging which may be referred to as a reactor support flange 280. The support flange 280 supports the weight of the reactor vessel 200 and internal components above the wet reactor well 115. In one embodiment, the support flange is structurally stiffened and reinforced by a plurality of lugs 281 which are spaced circumferentially apart around the reactor vessel and welded to both the reactor vessel and flange, as shown. Support flange contacts and engages the horizontal partition wall 114, which transfers the dead weight of the reactor vessel 200 to the containment vessel 110. The reactor vessel's radial and axial thermal expansion (i.e. a majority of growth being primarily downwards from the horizontal partition wall 114) as the reactor heats up during operation is unconstrained. However, the portion of the containment vessel 110 which projects above the partition wall 114 is free to grow upwards in unison with the upwards growth of the steam generating vessel 30 to minimize axial differential expansion between the steam generating vessel and reactor vessel. Because the reactor vessel and steam generating vessel are configured and structured to thermally grow in height at substantially the same rate when heated, this arrangement helps minimize potential thermal expansions stress in the primary coolant fluid coupling 273 between the reactor vessel and steam generating vessel. The support flange 280 is spaced vertically downwards on the reactor vessel shell 201 by a distance from the top head 202 of the reactor vessel 200 sufficient to allow a fluid connection to be made to the steam generating vessel 300 which is above the partition wall 114, as shown in FIGS. 1 and 2. When the reactor vessel 200 is mounted inside the containment vessel 110, the top head 202 of the reactor vessel and the primary coolant fluid coupling 273 (collectively formed by combined the inlet-outlet flow nozzle 270/271 and the inlet-outlet flow nozzle 371/370 of the steam generating vessel 300, shown in FIG. 4) are located above the reactor well 115. This provides a location for connection to the steam generator headers and for the engineered safety systems (e.g. control rods, etc.) to deal with various postulated accident scenarios. A majority of the reactor vessel shell 201, however, may be disposed below the partition wall 114 and immersed in the wet reactor well 115 as shown in FIG. 1. The bottom region of the reactor vessel 200 is restrained by a lateral seismic restraint system 260 (shown schematically in FIG. 1) that spans the space between the reactor shell 201 and the reactor well 115 inside surface of the cylindrical enclosure 116 to withstand seismic events. The seismic restraint design is configured to allow for free axial (i.e. longitudinal along vertical axis VA1) and diametrical thermal expansion of the reactor vessel 200. The reactor well 115 is flooded during power operations to provide defense-in-depth against a (hypothetical, non-mechanistic) accident that is assumed to produce a rapid rise in the enthalpy of the reactor's contents. Because the reactor is designed to prevent loss of core water by leaks or breaks and the reactor well is flooded, burn-through of the reactor vessel by molten fuel (corium) is not likely. Referring to FIGS. 3 and 4, the combined inlet-outlet flow nozzle 270/271 has two concentric hollow forgings including an outer inlet nozzle 270 and an inner outlet nozzle 271. The outlet nozzle 271 has one end welded to the reactor shroud 220 (internal to the reactor vessel shell 201) and an opposite end welded to the inlet nozzle 371 of the steam generating vessel 300. The inlet nozzle 270 has one end welded to the reactor vessel shell 201 and an opposite end welded to the outlet nozzle 370 of the steam generating vessel 300. The flow isolation plate 227 helps ensure that the hot primary coolant water exiting the reactor vessel cannot flow back into the annulus 221. In the present embodiment, the outlet nozzle 271 of the reactor vessel and the inlet nozzle 371 of the steam generating vessel each have a smaller diameter than the inlet nozzle 270 of the reactor vessel and the outlet nozzle 270 of the steam generating vessel. The combined inlet-outlet flow nozzle 270/271 is located above the partition wall 114 of the containment vessel 110. The inlet nozzle 371 and the outlet nozzle 370 of the steam generating vessel 300 collectively define a mating concentrically arranged combined inlet/outlet nozzle 371/370 for the steam generating vessel. In order to avoid long loops of large piping in the reactor primary coolant system which creates the potential for a “large break” LOCA event, both the combined inlet-outlet flow nozzle 270/271 of the reactor vessel 200 and the combined inlet/outlet nozzle 371/370 for the steam generating vessel are intentionally very closely coupled to the shells of their respective vessels having a minimal radial projection beyond the shells. The design of the top of the reactor vessel, with the flanged connection between the head and the shell of the reactor vessel, helps to minimize this radial projection beyond the shell. This is accomplished by reducing the extent to which the flanges extend out from the shell, as compared to the prior art. In addition, cost advantages may be realized in having the inlet-outlet flow nozzle 270/271 shortened, in that different manufacturing techniques may be used to create the shorter inlet-outlet flow nozzle 270/271 as compared to if a longer flow nozzle is required. This permits the reactor vessel 200 to be directly coupled to the steam generating vessel 300 via the inlet/outlet nozzles as shown in FIGS. 1 and 2. As shown in FIG. 3A, the combined inlet-outlet flow nozzle 270/271 of the reactor vessel preferably protrudes radially beyond the shell 201 by a distance that is no more than the radial projection of the support flange 280. The total length of the inlet/outlet nozzle connection between the reactor vessel 200 and steam generating vessel 300 in certain embodiment is less than or equal to the diameter of the reactor vessel 200, and/or the steam generating vessel 300 to eliminate long runs of large coolant piping between the reactor and steam generating vessels. In one embodiment, the nozzle connections between the reactor vessel 200 and the steam generating vessel 300 is straight without any elbows or bends. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims. |
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abstract | The present invention is directed to a scanning apparatus and method for processing a substrate, wherein the scanning apparatus comprises a base portion and a rotary subsystem. The rotary subsystem comprises a first link comprising a first joint, wherein the first link is rotatably coupled to the base portion by the first joint, and a second link comprising a second joint, wherein the second link is rotatably coupled to the first link by the second joint. The first joint and the second joint are spaced a predetermined distance from one another. The second link further comprising an end effector whereon the substrate resides, and wherein the end effector is operably coupled to the second link. The end effector is further spaced from the second joint by the predetermined distance, wherein a rotation of the first link and second link in a respective first direction and second direction is operable to linearly oscillate the end effector along a linear first scan path, and wherein the rotational velocity of the first link and second link does not cross zero. |
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claims | 1. A method for non-intrusively monitoring the gaseous contents of a sealed container comprising: transmitting an ultrasonic transmitted signal through a wall of the sealed container into the gaseous contents of the container from a transmitter; receiving an ultrasound signal from within the container using a receiver; and analyzing the changes in the received ultrasound signal so as to deduce the level of one or more components of the gaseous contents of the container due to the way in which the transmitted signal is modified into the received ultrasound signal by the gaseous contents of the container, without having to unseal the container, and wherein the deduced level of one or more components of the gaseous contents of the container is corrected for temperature by performing the method at a number of frequencies; providing a structural wall so as to isolate the gaseous contents of the container from the transmitter and the receiver, the structural wall being separate from the the one or more components of the gaseous contents of the container arising from a leakage of gas into the gaseous contents of the container, and wherein the transmitter and/or receiver can be removed without having to unseal the container. 2. A method according to claim 1 wherein the level of one or more components being deduced is the level of air ma positive helium atmosphere of at least 1.1 atmospheres, air arising from a leakage of the atmosphere surrounding the sealed container into the container atmosphere. claim 1 3. A method according to claim 1 wherein the sealed container contains nuclear fuel rods. claim 1 4. A method according to claim 3 wherein level of one or more components being deduced is the level of the act of transmitting includes the gaseous contents xenon and/or krypton in air and helium, the xenon and/or krypton arising from the leakage of fission product gases from the nuclear fuel rods within the sealed container. claim 3 5. A method according to claim 3 further comprising deducing whether the gaseous contents includes oxygen and/or fission product gasses from the fuel rods. claim 3 6. A method for non-intrusively monitoring the gaseous contents of a sealed container comprising: transmitting an ultrasonic transmitted signal through a wall of a monitoring housing, from a transmitter provided in the monitoring housing, and into a monitoring location, the monitoring location being in sealed fluid communication with the gaseous contents of the sealed container; receiving an ultrasound signal from within the monitoring location using a receiver provided in the monitoring housing; and analyzing the changes in the received ultrasound signal so as to deduce the level of one or more components of the gaseous contents of the container due to the way in which the transmitted signal is modified into the received ultrasound signal by the gaseous contents of the container, without having to unseal the container or the monitoring housing; providing a structural wall as a part of the monitoring housing so as to isolate the gaseous contents of the monitoring location and container from the transmitter and the receiver, the transmitter and receiver being separate from the monitoring housing, the one or more components arising from a leakage of gas into the gaseous contents of the container. 7. A method according to claim 6 in which the monitoring location is connected to the gaseous contents within the container via a passageway. claim 6 8. A method according to claim 6 in which the passageway is provided with one or more dog legs. claim 6 9. A method according to claim 7 in which the passageway passes through the lid of the container. claim 7 10. A method according to claim 6 , wherein the act of analysing includes measuring the sound velocity and attenuation of the transmitted signal and/or the received signal. claim 6 11. A method according to claim 10 further comprising the act of measuring the sound velocity and attenuation of the transmitted signal at more than one frequency. claim 10 12. A method according to claim 6 , wherein the act of analysing includes measuring the velocity of both a xe2x80x9cline of sightxe2x80x9d received signal and a reflected signal. claim 6 13. A method according to claim 6 wherein the act of analysing includes subjecting the received signal to a signal processing to extract the desired information, the signal processing involving Fast Fourier Transformation and/or chromatic based processing. claim 6 14. A method according to claim 13 in which chromatic based processing is employed, the signal processing involving the application of one or more Gaussian processors to the received signal, the processors being non-orthogonal. claim 13 15. An apparatus for non-intrusively monitoring the gaseous contents of a sealed container, the apparatus comprising: a transmitter for transmitting an ultrasonic signal through a wall of a sealed container to read, an ultrasonic signal through the wall of a monitoring housing of the sealed container container into the gaseous contents of the container; a receiver for receiving an ultrasound signal from within the container: an analyzer for analyzing changes in the received ultrasound signal so as to deduce the level of one or more components of the gaseous contents of the container due to the way in which the transmitted signal is modified into the received signal by the gaseous contents of the container, without having to unseal the containers a structural wall being provided so as to isolate the gaseous contents of the container from the transmitter and the receiver, the structural wall being separate from the transmitter and receiver, the one or more components arising from a leakage of gas into the gaseous contents of the container. 16. An apparatus according to claim 15 wherein the source of the transmitter is positioned outside the container, and the receiver is positioned outside the container. claim 15 17. An apparatus according to claim 16 , wherein the transmitter is positioned within the container. claim 16 18. An apparatus according to claim 17 , wherein the transmitting means is activated by a signal transmitted from outside the container. claim 17 19. An apparatus according to claim 15 , wherein the transmitter comprises a tuning fork or resonant cavity. claim 15 20. A method according to claim 6 in which the monitoring location is provided within a housing, the housing being provided with a volume of gas on the distal side of the monitoring location relative to the container, the volume of gas on the distal side having a greater extent, measured perpendicular to the passageway leading from the container to the monitoring location, than the extent of the monitoring location itself. claim 6 21. An apparatus according to claim 15 , in which the transmitter and/or receiver are provided in bores in the housing. claim 15 |
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abstract | A collimator assembly for use with radiation therapy systems, in particular stereotactic radiosurgery (“SRS”) systems, is provided. In general, the collimator assembly includes a collimator in which a plurality of concentric, conical slits is formed. Each conical slit is oriented along a different slit angle, such that radiation impinging on the top surface of the collimator is redirected along each conical slit towards a common target isocenter located at a distance away from the bottom surface of the collimator. The conical slits can be referred to as “Compton slits” because they are designed to increase the radiation output at the isocenter of the collimator by redirecting Compton scattered radiation towards the isocenter in the target region. The collimator assembly thus improves the efficiency of a radiation treatment system by utilizing Compton scattered radiation that would otherwise be lost. |
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description | An ideal fusion reactor solves the problem of anomalous transport for both ions and electrons. The anomalous transport of ions is avoided by magnetic confinement in a field reversed configuration (FRC) in such a way that the majority of the ions have large, non-adiabatic orbits, making them insensitive to short-wavelength fluctuations that cause anomalous transport of adiabatic ions. For electrons, the anomalous transport of energy is avoided by tuning the externally applied magnetic field to develop a strong electric field, which confines them electrostatically in a deep potential well. Moreover, the fusion fuel plasmas that can be used with the present confinement process and apparatus are not limited to neutronic fuels only, but also advantageously include advanced fuels. (For a discussion of advanced fuels, see R. Feldbacher and M. Heindler, Nuclear Instruments and Methods in Physics Research, A271(1988)JJ-64 (North Holland Amsterdam).) The solution to the problem of anomalous transport found herein makes use of a specific magnetic field configuration, which is the FRC. In particular, the existence of a region in a FRC where the magnetic field vanishes makes it possible to have a plasma comprising a majority of non-adiabatic ions. Background Theory Before describing the system and apparatus in detail, it will be helpful to first review a few key concepts necessary to understand the concepts contained herein. A particle with electric charge q moving with velocity {right arrow over (v)} in a magnetic field {right arrow over (B)} experiences a force {right arrow over (F)}L given by F → L = q v → xc3x97 B → c . ( 1 ) The force {right arrow over (F)}L is called the Lorentz force. It, as well as all the formulas used in the present discussion, is given in the gaussian system of units. The direction of the Lorentz force depends on the sign of the electric charge q. The force is perpendicular to both velocity and magnetic field. FIG. 1A shows the Lorentz force 30 acting on a positive charge. The velocity of the particle is shown by the vector 32. The magnetic field is 34. Similarly, FIG. 1B shows the Lorentz force 30 acting on a negative charge. As explained, the Lorentz force is perpendicular to the velocity of a particle; thus, a magnetic field is unable to exert force in the direction of the particle""s velocity. It follows from Newton""s second law, {right arrow over (F)}=m{right arrow over (a)}, that a magnetic field is unable to accelerate a particle in the direction of its velocity. A magnetic field can only bend the orbit of a particle, but the magnitude of its velocity is not affected by a magnetic field. FIG. 2A shows the orbit of a positively charged particle in a constant magnetic field 34. The Lorentz force 30 in this case is constant in magnitude, and the orbit 36 of the particle forms a circle. This circular orbit 36 is called a Larmor orbit. The radius of the circular orbit 36 is called a gyroradius 38. Usually, the velocity of a particle has a component that is parallel to the magnetic field and a component that is perpendicular to the field. In such a case, the particle undergoes two simultaneous motions: a rotation around the magnetic field line and a translation along it. The combination of these two motions creates a helix that follows the magnetic field line 40. This is indicated in FIG. 2B. A particle in its Larmor orbit revolves around a magnetic field line. The number of radians traveled per unit time is the particle""s gyrofrequency, which is denoted by xcexa9 and given by Ω = q xe2x80x83 B mc , ( 2 ) where m is the mass of the particle and c is the speed of light. The gyroradius aL of a charged particle is given by a L = v ⊥ Ω , ( 3 ) where xcexdxe2x8axa5 is the component of the velocity of the particle perpendicular to the magnetic field. {right arrow over (E)}xc3x97{right arrow over (B)} Drift and Gradient Drift Electric fields affect the orbits of charged particles, as shown in FIG. 3. In FIG. 3, the magnetic field 44 points toward the reader. The orbit of a positively charged ion due to the magnetic field 44 alone would be a circle 36; the same is true for an electron 42. In the presence of an electric field 46, however, when the ion moves in the direction of the electric field 46, its velocity increases. As can be appreciated, the ion is accelerated by the force q{right arrow over (E)}. It can further be seen that, according to Eq. 3, the ion""s gyroradius will increase as its velocity does. As the ion is accelerated by the electric field 46, the magnetic field 44 bends the ion""s orbit. At a certain point the ion reverses direction and begins to move in a direction opposite to the electric field 46. When this happens, the ion is decelerated, and its gyroradius therefore decreases. The ion""s gyroradius thus increases and decreases in alternation, which gives rise to a sideways drift of the ion orbit 48 in the direction 50 as shown in FIG. 3. This motion is called {right arrow over (E)}xc3x97{right arrow over (A)} drift. Similarly, electron orbits 52 drift in the same direction 50. A similar drift can be caused by a gradient of the magnetic field 44 as illustrated in FIG. 4. In FIG. 4, the magnetic field 44 points towards the reader. The gradient of the magnetic field is in the direction 56. The increase of the magnetic field""s strength is depicted by the denser amount of dots in the figure. From Eqs. 2 and 3, it follows that the gyroradius is inversely proportional to the strength of the magnetic field. When an ion moves in the direction of increasing magnetic field its gyroradius will decrease, because the Lorentz force increases, and vice versa. The ion""s gyroradius thus decreases and increases in alternation, which gives rise to a sideways drift of the ion orbit 58 in the direction 60. This motion is called gradient drift. Electron orbits 62 drift in the opposite direction 64. Adiabatic and Non-Adiabatic Particles Most plasma comprises adiabatic particles. An adiabatic particle tightly follows the magnetic field lines and has a small gyroradius. FIG. 5 shows a particle orbit 66 of an adiabatic particle that follows tightly a magnetic field line 68. The magnetic field lines 68 depicted are those of a Tokamak. A non-adiabatic particle has a large gyroradius. It does not follow the magnetic field lines and is usually energetic. There exist other plasmas that comprise non-adiabatic particles. FIG. 6 illustrates a non-adiabatic plasma for the case of a betatron. The pole pieces 70 generate a magnetic field 72. As FIG. 6 illustrates, the particle orbits 74 do not follow the magnetic field lines 72. Radiation in Plasmas A moving charged particle radiates electromagnetic waves. The power radiated by the particle is proportional to the square of the charge. The charge of an ion is Ze, where e is the electron charge and Z is the atomic number. Therefore, for each ion there will be Z free electrons that will radiate. The total power radiated by these Z electrons is proportional to the cube of the atomic number (Z3). Charged Particles in a FRC FIG. 8 shows the magnetic field of a FRC. The system has cylindrical symmetry with respect to its axis 78. In the FRC, there are two regions of magnetic field lines: open 80 and closed 82. The surface dividing the two regions is called the separatrix 84. The FRC forms a cylindrical null surface 86 in which the magnetic field vanishes. In the central part 88 of the FRC the magnetic field does not change appreciably in the axial direction. At the ends 90, the magnetic field does change appreciably in the axial direction. The magnetic field along the center axis 78 reverses direction in the FRC, which gives rise to the term xe2x80x9cReversedxe2x80x9d in Field Reversed Configuration (FRC). In FIG. 9A, the magnetic field outside of the null surface 94 is in the direction 96. The magnetic field inside the null surface is in the direction 98. If an ion moves in the direction 100, the Lorentz force 30 acting on it points towards the null surface 94. This is easily appreciated by applying the right-hand rule. For particles moving in the direction 102, called diamagnetic, the Lorentz force always points toward the null surface 94. This phenomenon gives rise to a particle orbit called betatron orbit, to be described below. FIG. 9B shows an ion moving in the direction 104, called counterdiamagnetic. The Lorentz force in this case points away from the null surface 94. This phenomenon gives rise to a type of orbit called a drift orbit, to be described below. The diamagnetic direction for ions is counterdiamagnetic for electrons, and vice versa. FIG. 10 shows a ring or annular layer of plasma 106 rotating in the ions"" diamagnetic direction 102. The ring 106 is located around the null surface 86. The magnetic field 108 created by the annular plasma layer 106, in combination with an externally applied magnetic field 110, forms a magnetic field having the topology of a FRC (The topology is shown in FIG. 8). The ion beam that forms the plasma layer 106 has a temperature; therefore, the velocities of the ions form a Maxwell distribution in a frame rotating at the average angular velocity of the ion beam. Collisions between ions of different velocities lead to fusion reactions. For this reason, the plasma beam layer 106 is called a colliding beam system. FIG. 11 shows the main type of ion orbits in a colliding beam system, called a betatron orbit 112. A betatron orbit 112 can be expressed as a sine wave centered on the null circle 114. As explained above, the magnetic field on the null circle 114 vanishes. The plane of the orbit 112 is perpendicular to the axis 78 of the FRC. Ions in this orbit 112 move in their diamagnetic direction 102 from a starting point 116. An ion in a betatron orbit has two motions: an oscillation in the radial direction (perpendicular to the null circle 114), and a translation along the null circle 114. FIG. 12A is a graph of the magnetic field 118 in a FRC. The field 118 is derived using a one-dimensional equilibrium model, to be discussed below in conjunction with the theory of the invention. The horizontal axis of the graph represents the distance in centimeters from the FRC axis 78. The magnetic field is in kilogauss. As the graph depicts, the magnetic field 118 vanishes at the null circle radius 120. As shown in FIG. 12B, a particle moving near the null circle will see a gradient 126 of the magnetic field pointing away from the null surface 86. The magnetic field outside the null circle is 122, while the magnetic field inside the null circle is 124. The direction of the gradient drift is given by the cross product {right arrow over (B)}xc3x97∇B, where ∇B is the gradient of the magnetic field; thus, it can be appreciated by applying the right-hand rule that the direction of the gradient drift is in the counterdiamagnetic direction, whether the ion is outside or inside the null circle 128. FIG. 13A is a graph of the electric field 130 in a FRC. The field 130 is derived using a one-dimensional equilibrium model, to be discussed below in conjunction with the theory of the invention. The horizontal axis of the graph represents the distance in centimeters from the FRC axis 78. The electric field is in volts/cm. As the graph depicts, the electric field 130 vanishes close to the null circle radius 120. As shown if FIG. 13B, the electric field for ions is deconfining; it points away from the null surface 132, 134. The magnetic field, as before, is in the directions 122, 124. It can be appreciated by applying the right-hand rule that the direction of the {right arrow over (E)}xc3x97{right arrow over (B)} drift is in the diamagnetic direction, whether the ion is outside or inside the null surface 136. FIGS. 14A and 14B show another type of common orbit in a FRC, called a drift orbit 138. Drift orbits 138 can be outside of the null surface, as shown in FIG. 14A, or inside it, as shown in FIG. 14B. Drift orbits 138 rotate in the diamagnetic direction if the {right arrow over (E)}xc3x97{right arrow over (B)} drift dominates or in the counterdiamagnetic direction if the gradient drift dominates. The drift orbits 138 shown in FIGS. 14A and 14B rotate in the diamagnetic direction 102 from starting point 116. A drift orbit, as shown in FIG. 14C, can be thought of as a small circle rolling over a relatively bigger circle. The small circle 142 spins around its axis in the sense 144. It also rolls over the big circle 146 in the direction 102. The point 140 will trace in space a path similar to 138. FIGS. 15A and 15B show the direction of the Lorentz force at the ends of a FRC. In FIG. 15A, an ion is shown moving in the diamagnetic direction 102 with a velocity 148 in a magnetic field 150. It can be appreciated by applying the right-hand rule that the Lorentz force 152 tends to push the ion back into the region of closed field lines. In this case, therefore, the Lorentz force 152 is confining for the ions. In FIG. 15B, an ion is shown moving in the counterdiamagnetic direction with a velocity 148 in a magnetic field 150. It can be appreciated by applying the right-hand rule that the Lorentz force 152 tends to push the ion into the region of open field lines. In this case, therefore, the Lorentz force 152 is deconfining for the ions. Magnetic and Electrostatic Confinement in a FRC A plasma layer 106 (see FIG. 10) can be formed in a FRC by injecting energetic ion beams around the null surface 86 in the diamagnetic direction 102 of ions. (A detailed discussion of different methods of forming the FRC and plasma ring follows below.) In the circulating plasma layer 106, most of the ions have betatron orbits 112 (see FIG. 11), are energetic, and are non-adiabatic; thus, they are insensitive to short-wavelength fluctuations that cause anomalous transport. While studying a plasma layer 106 in equilibrium conditions as described above, it was discovered that the conservation of momentum imposes a relation between the angular velocity of ions xcfx89i and the angular velocity of electrons xcfx89e. (The derivation of this relation is given below in conjunction with the theory of the invention.) The relation is ω e = ω i [ 1 - ω i Ω 0 ] , where xe2x80x83 Ω 0 = Ze xe2x80x83 B 0 m i c . ( 4 ) In Eq. 4, Z is the ion atomic number, mi is the ion mass, e is the electron charge, B0 is the magnitude of the applied magnetic field, and c is the speed of light. There are three free parameters in this relation: the applied magnetic field B0, the electron angular velocity xcfx89e, and the ion angular velocity xcfx89i. If two of them are known, the third can be determined from Eq. 4. Because the plasma layer 106 is formed by injecting ion beams into the FRC, the angular velocity of ions xcfx89i is determined by the injection kinetic energy of the beam Wi, which is given by W i = 1 2 m i V i 2 = 1 2 ( m i ( ω i r o ) ) 2 . Here, Vi=xcfx89ir0, where Vi is the injection velocity of ions, xcfx89i is the cyclotron frequency of ions, and r0 is the radius of the null surface 86. The kinetic energy of electrons in the beam has been ignored because the electron mass me is much smaller than the ion mass mi. For a fixed injection velocity of the beam (fixed xcfx89i), the applied magnetic field B0 can be tuned so that different values of xcfx89e are obtainable. As will be shown, tuning the external magnetic field B0 also gives rise to different values of the electrostatic field inside the plasma layer. This feature of the invention is illustrated in FIGS. 16A and 16B. FIG. 16A shows three plots of the electric field (in volts/cm) obtained for the same injection velocity, xcfx89i=1.35xc3x97107sxe2x88x921, but for three different values of the applied magnetic field B0: The values of xcfx89e in the table above were determined according to Eq. 4. One can appreciate that xcfx89e greater than 0 means that xcexa90 greater than xcfx89i in Eq. 4, so that electrons rotate in their counterdiamagnetic direction. FIG. 16B shows the electric potential (in volts) for the same set of values of B0 and xcfx89e. The horizontal axis, in FIGS. 16A and 16B, represents the distance from the FRC axis 78, shown in the graph in centimeters. The analytic expressions of the electric field and the electric potential are given below in conjunction with the theory of the invention. These expressions depend strongly on xcfx89e. The above results can be explained on simple physical grounds. When the ions rotate in the diamagnetic direction, the ions are confined magnetically by the Lorentz force. This was shown in FIG. 9A. For electrons, rotating in the same direction as the ions, the Lorentz force is in the opposite direction, so that electrons would not be confined. The electrons leave the plasma and, as a result, a surplus of positive charge is created. This sets up an electric field that prevents other electrons from leaving the plasma. The direction and the magnitude of this electric field, in equilibrium, is determined by the conservation of momentum. The relevant mathematical details are given below in conjunction with the theory of the invention. The electrostatic field plays an essential role on the transport of both electrons and ions. Accordingly, an important aspect of this invention is that a strong electrostatic field is created inside the plasma layer 106, the magnitude of this electrostatic field is controlled by the value of the applied magnetic field B0 which can be easily adjusted. As explained, the electrostatic field is confining for electrons if xcfx89e greater than 0. As shown in FIG. 16B, the depth of the well can be increased by tuning the applied magnetic field B0. Except for a very narrow region near the null circle, the electrons always have a small gyroradius. Therefore, electrons respond to short-wavelength fluctuations with an anomalously fast diffusion rate. This diffusion, in fact, helps maintain the potential well once the fusion reaction occurs. The fusion product ions, being of much higher energy, leave the plasma. To maintain charge quasi-neutrality, the fusion products must pull electrons out of the plasma with them, mainly taking the electrons from the surface of the plasma layer. The density of electrons at the surface of the plasma is very low, and the electrons that leave the plasma with the fusion products must be replaced; otherwise, the potential well would disappear. FIG. 17 shows a Maxwellian distribution 162 of electrons. Only very energetic electrons from the tail 160 of the Maxwell distribution can reach the surface of the plasma and leave with fusion ions. The tail 160 of the distribution 162 is thus continuously created by electronxe2x80x94electron collisions in the region of high density near the null surface. The energetic electrons still have a small gyroradius, so that anomalous diffusion permits them to reach the surface fast enough to accommodate the departing fusion product ions. The energetic electrons lose their energy ascending the potential well and leave with very little energy. Although the electrons can cross the magnetic field rapidly, due to anomalous transport, anomalous energy losses tend to be avoided because little energy is transported. Another consequence of the potential well is a strong cooling mechanism for electrons that is similar to evaporative cooling. For example, for water to evaporate, it must be supplied the latent heat of vaporization. This heat is supplied by the remaining liquid water and the surrounding medium, which then thermalize rapidly to a lower temperature faster than the heat transport processes can replace the energy. Similarly, for electrons, the potential well depth is equivalent to water""s latent heat of vaporization. The electrons supply the energy required to ascend the potential well by the thermalization process that re-supplies the energy of the Maxwell tail so that the electrons can escape. The thermalization process thus results in a lower electron temperature, as it is much faster than any heating process. Because of the mass difference between electrons and protons, the energy transfer time from protons is about 1800 times less than the electron thermalization time. This cooling mechanism also reduces the radiation loss of electrons. This is particularly important for advanced fuels, where radiation losses are enhanced by fuel ions with atomic number Z greater than 1. The electrostatic field also affects ion transport. The majority of particle orbits in the plasma layer 106 are betatron orbits 112. Large-angle collisions, that is, collisions with scattering angles between 90xc2x0 and 180xc2x0, can change a betatron orbit to a drift orbit. As described above, the direction of rotation of the drift orbit is determined by a competition between the {right arrow over (E)}xc3x97{right arrow over (B)} drift and the gradient drift. If the {right arrow over (E)}xc3x97{right arrow over (B)} drift dominates, the drift orbit rotates in the diamagnetic direction. If the gradient drift dominates, the drift orbit rotates in the counterdiamagnetic direction. This is shown in FIGS. 18A and 18B. FIG. 18A shows a transition from a betatron orbit to a drift orbit due to a 180xc2x0 collision, which occurs at the point 172. The drift orbit continues to rotate in the diamagnetic direction because the {right arrow over (E)}xc3x97{right arrow over (B)} drift dominates. FIG. 18B shows another 180xc2x0 collision, but in this case the electrostatic field is weak and the gradient drift dominates. The drift orbit thus rotates in the counterdiamagnetic direction. The direction of rotation of the drift orbit determines whether it is confined or not. A particle moving in a drift orbit will also have a velocity parallel to the FRC axis. The time it takes the particle to go from one end of the FRC to the other, as a result of its parallel motion, is called transit time; thus, the drift orbits reach an end of the FRC in a time of the order of the transit time. As shown in connection with FIG. 15A, the Lorentz force at the ends is confining only for drift orbits rotating in the diamagnetic direction. After a transit time, therefore, ions in drift orbits rotating in the counterdiamagnetic direction are lost. This phenomenon accounts for a loss mechanism for ions, which is expected to have existed in all FRC experiments. In fact, in these experiments, the ions carried half of the current and the electrons carried the other half. In these conditions the electric field inside the plasma was negligible, and the gradient drift always dominated the {right arrow over (E)}xc3x97{right arrow over (B)} drift. Hence, all the drift orbits produced by large-angle collisions were lost after a transit time. These experiments reported ion diffusion rates that were faster than those predicted by classical diffusion estimates. If there is a strong electrostatic field, the {right arrow over (E)}xc3x97{right arrow over (B)} drift dominates the gradient drift, and the drift orbits rotate in the diamagnetic direction. This was shown above in connection with FIG. 18A. When these orbits reach the ends of the FRC, they are reflected back into the region of closed field lines by the Lorentz force; thus, they remain confined in the system. The electrostatic fields in the colliding beam system may be strong enough, so that the {right arrow over (E)}xc3x97{right arrow over (B)} drift dominates the gradient drift. Thus, the electrostatic field of the system would avoid ion transport by eliminating this ion loss mechanism, which is similar to a loss cone in a mirror device. Another aspect of ion diffusion can be appreciated by considering the effect of small-angle, electron-ion collisions on betatron orbits. FIG. 19A shows a betatron orbit 112; FIG. 19B shows the same orbit 112 when small-angle electron-ion collisions are considered 174; FIG. 19C shows the orbit of FIG. 19B followed for a time that is longer by a factor often 176; and FIG. 19D shows the orbit of FIG. 19B followed for a time longer by a factor of twenty 178. It can be seen that the topology of betatron orbits does not change due to small-angle, electron-ion collisions; however, the amplitude of their radial oscillations grows with time. In fact, the orbits shown in FIGS. 19A to 19D fatten out with time, which indicates classical diffusion. Theory of the Invention For the purpose of modeling the invention, a one-dimensional equilibrium model for the colliding beam system is used, as shown in FIG. 10. The results described above were drawn from this model. This model shows how to derive equilibrium expressions for the particle densities, the magnetic field, the electric field, and the electric potential. The equilibrium model presented herein is valid for a plasma fuel with one type of ions (e.g., in a D-D reaction) or multiple types of ions (e.g., D-T, D-He3, and p-B11). Vlasov-Maxwell Equations Equilibrium solutions for the particle density and the electromagnetic fields in a FRC are obtained by solving self-consistently the Vlasov-Maxwell equations: ∂ f j ∂ t + ( v → · ∇ ) f j + e j m j [ E → + v → c xc3x97 B → ] · ∇ v f j = 0 ( 5 ) ∇ xc3x97 E → = - 1 c ∂ B → ∂ t ( 6 ) ∇ xc3x97 B → = 4 π c ∑ j e j ∫ v → f j ⅆ v → + 1 c ∂ E → ∂ t ( 7 ) ∇ · E → = 4 π ∑ j e j ∫ f j ⅆ v → ( 8 ) xe2x80x83∇xc2x7{right arrow over (B)}=0,xe2x80x83xe2x80x83(9) where j=e, i and i=1, 2, . . . for electrons and each species of ions. In equilibrium, all physical quantities are independent of time (i.e., ∂/∂t=0). To solve the Vlasov-Maxwell equations, the following assumptions and approximations are made: (a) All the equilibrium properties are independent of axial position z (i.e., ∂/∂z=0). This corresponds to considering a plasma with an infinite extension in the axial direction; thus, the model is valid only for the central part 88 of a FRC. (b) The system has cylindrical symmetry. Hence, all equilibrium properties do not depend on xcex8(i.e., ∂/∂xcex8=0). (c) The Gauss law, Eq. 8, is replaced with the quasi-neutrality condition: xcexa3jnjej=0. By assuming infinite axial extent of the FRC and cylindrical symmetry, all the equilibrium properties will depend only on the radial coordinate r. For this reason, the equilibrium model discussed herein is called one-dimensional. With these assumptions and approximations, the Vlasov-Maxwell equations reduce to: ( v → · ∇ ) f j + e j m j E → · ∇ v f j + e j m j c [ v → xc3x97 B → ] · ∇ v f j = 0 ( 10 ) ∇ xc3x97 B → = 4 π c ∑ j e j ∫ v → f j ⅆ v → ( 11 ) ∑ α n j e j = 0. ( 12 ) Rigid Rotor Distributions To solve Eqs. 10 through 12, distribution functions must be chosen that adequately describe the rotating beams of electrons and ions in a FRC. A reasonable choice for this purpose are the so-called rigid rotor distributions, which are Maxwellian distributions in a uniformly rotating frame of reference. Rigid rotor distributions are functions of the constants of motion: f j ( r , v → ) = ( m j 2 π xe2x80x83 T j ) 3 2 n j ( 0 ) exp [ - ϵ j - ω j P j T j ] , ( 13 ) where mj is particle mass, {right arrow over (v)} is velocity, Tj is temperature, nj(0) is density at r=0, and xcfx89j is a constant. The constants of the motion are ϵ j = m j 2 v 2 + e j Φ xe2x80x83 ( for xe2x80x83 energy ) xe2x80x83 and P j = m j ( xv y - yv x ) + e j c Ψ xe2x80x83 ( for xe2x80x83 canonical xe2x80x83 angular xe2x80x83 momentum ) , where "PHgr" is the electrostatic potential and xcexa8 is the flux function. The electromagnetic fields are E r = - ∂ Φ ∂ r xe2x80x83 ( electric xe2x80x83 field ) xe2x80x83 and B z = 1 r ∂ Ψ ∂ r xe2x80x83 ( magnetic xe2x80x83 field ) . Substituting the expressions for energy and canonical angular momentum into Eq. 13 yields f j ( r , v → ) = ( m j 2 π xe2x80x83 T j ) 3 2 n j ( r ) exp { - m j 2 xe2x80x83 T j "LeftBracketingBar" v → - ω → j xc3x97 r → "RightBracketingBar" 2 } , ( 14 ) where |{right arrow over (v)}xe2x88x92{right arrow over (xcfx89)}jxc3x97{right arrow over (r)}|2=(vx+yxcfx89j)2+(vyxe2x88x92xxcfx89j)2+vz2 and n j ( r ) = n j ( 0 ) exp { - 1 T j [ e j ( Φ - ω j c Ψ ) - m j 2 ω j 2 r 2 ] } . ( 15 ) That the mean velocity in Eq. 14 is a uniformly rotating vector gives rise to the name rigid rotor. One of skill in the art can appreciate that the choice of rigid rotor distributions for describing electrons and ions in a FRC is justified because the only solutions that satisfy Vlasov""s equation (Eq. 10) are rigid rotor distributions (e.g., Eq. 14). A proof of this assertion follows: Proof We require that the solution of Vlasov""s equation (Eq. 10) be in the form of a drifted Maxwellian: f j ( r → , v → ) = ( m j 2 π xe2x80x83 T j ( r ) ) 3 2 n j ( r ) exp [ - m α 2 xe2x80x83 T j ( r ) ( v → - u → j ( r ) ) 2 ] , ( 16 ) i.e., a Maxwellian with particle density nj(r), temperature Tj(r), and mean velocity uj(r) that are arbitrary functions of position. Substituting Eq. 16 into the Vlasov""s equation (Eq. 10) shows that (a) the temperatures Tj(r) must be constants; (b) the mean velocities {right arrow over (u)}j(r) must be uniformly rotating vectors; and (c) the particle densities nj(r) must be of the form of Eq. 15. Substituting Eq. 16 into Eq. 10 yields a third-order polynomial equation in {right arrow over (v)}: v → · ∇ ( ln xe2x80x83 n j ) + m j ( v → - u → j ) T j · ( v → · ∇ ) u → j + m j ( v → - u → j ) 2 2 T j 2 ( v → · ∇ ) T j … … + e j T j E → · ( v → - u → j ) - e j T j c [ v → xc3x97 B → ] · ( v → - u → j ) = 0. Grouping terms of like order in {right arrow over (v)} yields m j 2 xe2x80x83 T j 2 "LeftBracketingBar" v → "RightBracketingBar" 2 ( v → · ∇ T j ) … … + m j T j ( v → · ∇ u → j · v → ) - m j T j 2 ( v → · u → j ) ( v → · ∇ T j ) … … + v → · ∇ ( ln xe2x80x83 n j ) + m j 2 T j 2 "LeftBracketingBar" u → j "RightBracketingBar" 2 ( v → · ∇ T j ) - m j T j ( v → · ∇ u → j · u → j ) - e j T j v → · E → + e j cT j ( v → xc3x97 B → ) · u → j … … + e j T j E → · u → j = 0. For this polynomial equation to hold for all {right arrow over (v)}, the coefficient of each power of {right arrow over (v)} must vanish. The third-order equation yields Tj(r)=constant. The second-order equation gives v → · ∇ u → j · v → = xe2x80x83 ( v x v y v z ) ( ∂ u x ∂ x ∂ u y ∂ x ∂ u z ∂ x ∂ u x ∂ y ∂ u y ∂ y ∂ u z ∂ y ∂ u x ∂ z ∂ u y ∂ z ∂ u z ∂ z ) ( v x v y v z ) = xe2x80x83 v x 2 ∂ u x ∂ x + v y 2 ∂ u y ∂ y + v z 2 ∂ u z ∂ z + v x v y ( ∂ u y ∂ x + ∂ u x ∂ y ) … xe2x80x83 … + v x v z ( ∂ u z ∂ x + ∂ u x ∂ z ) + v y v z ( ∂ u z ∂ y + ∂ u y ∂ z ) = 0. For this to hold for all {right arrow over (v)}, we must satisfy ∂ u x ∂ x = ∂ u y ∂ y = ∂ u z ∂ z = 0 xe2x80x83 and "IndentingNewLine" ( ∂ u y ∂ x + ∂ u x ∂ y ) = ( ∂ u z ∂ x + ∂ u x ∂ z ) = ( ∂ u z ∂ y + ∂ u y ∂ z ) = 0 , which is solved generally by {right arrow over (u)}j({right arrow over (r)})=({right arrow over (xcfx89)}jxc3x97{right arrow over (r)})+{right arrow over (u)}0jxe2x80x83xe2x80x83(17) In cylindrical coordinates, take {right arrow over (u)}0j=0 and {right arrow over (xcfx89)}j=xcfx89j{circumflex over (z)}, which corresponds to injection perpendicular to a magnetic field in the {circumflex over (z)} direction. Then, {right arrow over (u)}j({right arrow over (r)})=xcfx89jr{circumflex over (xcex8)}. The zero order equation indicates that the electric field must be in the radial direction, i.e., {right arrow over (E)}=Er{circumflex over (r)}. The first-order equation is now given by v → · ∇ ( ln xe2x80x83 n j ) - m j T j ( v → · ∇ u → j · u → j ) - e j T j v → · E → + e j c xe2x80x83 T j ( v → xc3x97 B → ) · u → j = 0. ( 18 ) The second term in Eq. 18 can be rewritten with ∇ u → j · u → j = ( ∂ u r ∂ r ∂ u θ ∂ r ∂ u z ∂ r 1 r ∂ u r ∂ θ 1 r ∂ u θ ∂ θ 1 r ∂ u z ∂ θ ∂ u r ∂ z ∂ u θ ∂ z ∂ u z ∂ z ) ( u r u θ u z ) = ( 0 ω j 0 0 0 0 0 0 0 ) ( 0 ω j r 0 ) = ω j 2 r r ^ . ( 19 ) The fourth term in Eq. 18 can be rewritten with ( v → xc3x97 B → ) · u → j = xe2x80x83 v → · ( B → xc3x97 u → j ) = xe2x80x83 v → · ( ( ∇ xc3x97 A → ) xc3x97 u → j ) = xe2x80x83 v → · [ ( 1 r ∂ ∂ r ( r xe2x80x83 A θ ) z ^ ) xc3x97 ( - ω j r xe2x80x83 θ ^ ) ] = xe2x80x83 v → · ω j ∂ ∂ r ( r xe2x80x83 A θ ) r ^ ( 20 ) Using Eqs. 19 and 20, the first-order Eq. 18 becomes ∂ ∂ r ( ln xe2x80x83 n j ) - m j T j ω j 2 r - e j T j E r + e j ω j c xe2x80x83 T j ∂ ∂ r ( r xe2x80x83 A θ ( r ) ) = 0. The solution of this equation is n j ( r ) = n j ( 0 ) exp [ m j ω j 2 r 2 2 T j - e j Φ ( r ) T j - e j ω j r xe2x80x83 A θ ( r ) c xe2x80x83 T j ] , ( 21 ) where Er=xe2x88x92d"PHgr"/dr and nj(0) is given by n j ( 0 ) = n j0 exp [ - m j ω j 2 r 0 2 2 T j + e j Φ ( r 0 ) T j + e j ω j r 0 A θ ( r 0 ) c xe2x80x83 T j ] . ( 22 ) Here, nj0 is the peak density at r0. Solution of Vlasov-Maxwell Equations Now that it has been proved that it is appropriate to describe ions and electrons by rigid rotor distributions, the Vlasov""s equation (Eq. 10) is replaced by its first-order moments, i.e., - n j m j r xe2x80x83 ω j 2 = n j e j [ E r + r xe2x80x83 ω j c B z ] - T j ⅆ n j ⅆ r , ( 23 ) which are conservation of momentum equations. The system of equations to obtain equilibrium solutions reduces to: - n j m j r xe2x80x83 ω j 2 = n j e j [ E r + r xe2x80x83 ω j c B z ] - T j ⅆ n j ⅆ r xe2x80x83 j = e , i = 1 , 2 , … ( 24 ) - ∂ ∂ r 1 r ∂ Ψ ∂ r = - ∂ B z ∂ r = 4 π c j θ = 4 π c r ∑ j n j e j ω j ( 25 ) ∑ j n j e j ≅ 0. ( 26 ) Solution for Plasma with One Type of Ion Consider first the case of one type of ion fully stripped. The electric charges are given by ej=xe2x88x92e,Ze. Solving Eq. 24 for Er with the electron equation yields E r = m e r xe2x80x83 ω e 2 - r xe2x80x83 ω e c B z - T e e xe2x80x83 n e ⅆ n e ⅆ r , ( 27 ) and eliminating Er from the ion equation yields 1 r xe2x80x83 ⅆ log xe2x80x83 n i ⅆ r = Z i e c ( ω i - ω e ) T i B z - Z 2 T e T i 1 r xe2x80x83 ⅆ log xe2x80x83 n e ⅆ r + m i ω i 2 T i + m xe2x80x83 Z i ω e 2 T i . ( 28 ) Differentiating Eq. 28 with respect to r and substituting Eq. 25 for dBz/dr yields - ⅆ B z ⅆ r = 4 π c n e e xe2x80x83 r ( ω i - ω e ) xe2x80x83 and xe2x80x83 Z i n i = n e , with Te=Ti=constant, and xcfx89i, xcfx89e, constants, obtaining 1 r ⅆ ⅆ r 1 r xe2x80x83 ⅆ log xe2x80x83 n i ⅆ r = 4 π xe2x80x83 n e Z i e 2 T i ( ω i - ω e ) 2 c 2 - Z i T e T i 1 r ⅆ ⅆ r 1 r ⅆ log xe2x80x83 n e ⅆ r . ( 29 ) The new variable "xgr" is introduced: ξ = r 2 2 r 0 2 ⇒ 1 r ⅆ ⅆ r 1 r ⅆ ⅆ r = 1 r 0 4 d 2 d 2 ξ . ( 30 ) Eq. 29 can be expressed in terms of the new variable "xgr": d 2 log xe2x80x83 n i d 2 ξ = 4 π xe2x80x83 n e Z i e 2 r 0 4 T i ( ω i - ω e ) 2 c 2 - Z i T e T i d 2 log xe2x80x83 n e d 2 ξ . ( 31 ) Using the quasi-neutrality condition, n e = Z i n i ⇒ d 2 log xe2x80x83 n e d 2 ξ = d 2 log xe2x80x83 n i d 2 ξ ,
yields
d 2 log xe2x80x83 n i d 2 ξ = - r 0 4 ( T i + Z i T e ) 4 π xe2x80x83 Z i 2 e 2 c 2 ( ω i - ω e ) 2
n i = - r 0 4 ( T e + T i Z i ) 4 π xe2x80x83 n e0 e 2 c 2 ( ω i - ω e ) 2
n i n i0 = - 8 ( r 0 Δ xe2x80x83 r ) 2 n i n i0 . ( 32 ) Here is defined r 0 Δ xe2x80x83 r ≡ 2 2 { T e + T i Z i 4 π xe2x80x83 n e0 e 2 } 1 2 c "LeftBracketingBar" ω i - ω e "RightBracketingBar" , ( 33 ) where the meaning of xcex94r will become apparent soon. If Ni=nini0, where ni0 is the peak density at r=r0, Eq. 32 becomes d 2 log xe2x80x83 N i d 2 ξ = - 8 ( r 0 Δ xe2x80x83 r ) 2 N i . ( 34 ) Using another new variable, χ = 2 r 0 Δ xe2x80x83 r ξ , yields xe2x80x83 d 2 N i d 2 χ = - 2 N i , the solution to which is N i = 1 cos xe2x80x83 h 2 ( χ - χ 0 ) , where "khgr"0="khgr"(r0) because of the physical requirement that Ni(r0)=1. Finally, the ion density is given by n i = xe2x80x83 n i0 cos xe2x80x83 h 2 2 ( r 0 Δ xe2x80x83 r ) ( ξ - 1 2 ) = xe2x80x83 n i0 cos xe2x80x83 h 2 ( r 2 - r 0 2 r 0 Δ xe2x80x83 r ) . ( 35 ) The significance of r0 is that it is the location of peak density. Note that ni(0)=ni({square root over (2)}r0). With the ion density known, Bz can be calculated using Eq. 11, and Er can be calculated using Eq. 27. The electric and magnetic potentials are "PHgr"=xe2x88x92∫rxe2x80x2=0rxe2x80x2=0Er(rxe2x80x2)drxe2x80x2 and Φ = xe2x80x83 - ∫ r xe2x80x2 = 0 r xe2x80x2 = r E r ( r xe2x80x2 ) ⅆ r xe2x80x2 xe2x80x83 and A θ = xe2x80x83 1 r ∫ r xe2x80x2 = 0 r xe2x80x2 = r r xe2x80x2 B z ( r xe2x80x2 ) ⅆ r xe2x80x2 Ψ = xe2x80x83 r xe2x80x83 A θ xe2x80x83 ( flux xe2x80x83 function ) ( 36 ) Taking r={square root over (2)}r0 to be the radius at the wall (a choice that will become evident when the expression for the electric potential "PHgr"(r) is derived, showing that at r={square root over (2)}r0 the potential is zero, i.e., a conducting wall at ground), the line density is N e = xe2x80x83 Z i N i = xe2x80x83 ∫ r = 0 r = 2 r 0 n e0 2 π ⅆ r cos xe2x80x83 h 2 ( r 2 - r 0 2 r 0 Δ xe2x80x83 r ) = xe2x80x83 2 π xe2x80x83 n e0 r 0 Δ xe2x80x83 r xe2x80x83 tan xe2x80x83 h xe2x80x83 r 0 Δ xe2x80x83 r … … ≅ xe2x80x83 2 π xe2x80x83 n e0 r 0 Δ xe2x80x83 r xe2x80x83 ( because xe2x80x83 r 0 ⪢ Δ xe2x80x83 r ) ( 37 ) Thus, xcex94r represents an xe2x80x9ceffective thickness.xe2x80x9d In other words, for the purpose of line density, the plasma can be thought of as concentrated at the null circle in a ring of thickness xcex94r with constant density ne0. The magnetic field is B z ( r ) = B z ( 0 ) - 4 π c ∫ r xe2x80x2 = 0 r xe2x80x2 = r ⅆ r xe2x80x2 n e e xe2x80x83 r xe2x80x2 ( ω i - ω e ) . ( 38 ) The current due to the ion and electron beams is I θ = ∫ 0 2 r 0 j θ ⅆ r = N e e ( ω i - ω e ) 2 π xe2x80x83 j θ = n 0 e xe2x80x83 r ( ω i - ω e ) . ( 39 ) Using Eq. 39, the magnetic field can be written as B z ( r ) = xe2x80x83 B z ( 0 ) - 2 π c I θ - 2 π c I θ tan xe2x80x83 h xe2x80x83 r 2 - r 0 2 r 0 Δ xe2x80x83 r = xe2x80x83 - B 0 - 2 π c I θ tan xe2x80x83 h xe2x80x83 r 2 - r 0 2 r 0 Δ xe2x80x83 r . In xe2x80x83 Eq . xe2x80x83 40 , xe2x80x83 B z ( 0 ) = - B 0 + 2 π c I θ xe2x80x83 and xe2x80x83 B z ( 2 r 0 ) = - B 0 - 2 π c I θ . ( 40 ) If the plasma current Ixcex8 vanishes, the magnetic field is constant, as expected. These relations are illustrated in FIGS. 20A through 20C. FIG. 20A shows the external magnetic field {right arrow over (B)}0 180. FIG. 20B shows the magnetic field due to the ring of current 182, the magnetic field having a magnitude of (2xcfx80/c)Ixcex8. FIG. 20C shows field reversal 184 due to the overlapping of the two magnetic fields 180, 182. The magnetic field is B z ( r ) = xe2x80x83 - B 0 [ 1 + 2 π xe2x80x83 I θ c xe2x80x83 B 0 tan xe2x80x83 h xe2x80x83 r 2 - r 0 2 r 0 Δ xe2x80x83 r ] = xe2x80x83 - B 0 [ 1 + β tan xe2x80x83 h ( r 2 - r 0 2 r 0 Δ xe2x80x83 r ) ] , ( 41 ) using the following definition for xcex2: 2 π c I θ B 0 = N e e ( ω i - ω e ) c xe2x80x83 B 0 = 2 π c n e0 r 0 Δ xe2x80x83 re ( ω i - ω e ) B 0 …
… = 2 π c 2 2 [ T e + ( T i / Z i ) 4 π xe2x80x83 n e0 e 2 ] 1 2 c xe2x80x83 n e0 ω i - ω e e ( ω i - ω e ) B 0 …
… = [ 8 π ( n e0 T e + n i0 T i ) B 0 2 ] 1 2 ≡ β . ( 42 ) With an expression for the magnetic field, the electric potential and the magnetic flux can be calculated. From Eq. 27, E r = - r xe2x80x83 ω e c B z - T e e xe2x80x83 ⅆ ln xe2x80x83 n e ⅆ r + m e r xe2x80x83 ω e 2 = - ⅆ Φ ⅆ r ( 43 ) Integrating both sides of Eq. 28 with respect to r and using the definitions of electric potential and flux function, Φ ≡ - ∫ r xe2x80x2 = 0 r xe2x80x2 = r E r ⅆ r xe2x80x2 xe2x80x83 and xe2x80x83 Ψ ≡ ∫ r xe2x80x2 = 0 r xe2x80x2 = r B z ( r xe2x80x2 ) r xe2x80x2 ⅆ r xe2x80x2 ,
which xe2x80x83 yields ( 44 ) Φ = ω e e Ψ + T e e ln xe2x80x83 n e ( r ) n e ( 0 ) - m e r 2 ω e 2 2 . ( 45 ) Now, the magnetic flux can be calculated directly from the expression of the magnetic field (Eq. 41): Ψ = ∫ r xe2x80x2 = 0 r xe2x80x2 = r - B 0 [ 1 + β tan xe2x80x83 h xe2x80x83 r 2 - r 0 2 r 0 Δ xe2x80x83 r ] r xe2x80x2 ⅆ r xe2x80x2 …
… = - B o r 2 2 - B 0 β 2 r 0 Δ xe2x80x83 r [ log ( cos xe2x80x83 h xe2x80x83 r 2 - r 0 2 r 0 Δ xe2x80x83 r ) - log ( cos xe2x80x83 h xe2x80x83 r o Δ xe2x80x83 r ) ] …
… = - B 0 r 2 2 + B 0 β r 0 Δ xe2x80x83 r 4 log xe2x80x83 n e ( r ) n e ( 0 ) . ( 46 ) Substituting Eq. 46 into Eq. 45 yields Φ = ω e c B 0 β r 0 Δ xe2x80x83 r 4 log xe2x80x83 n e ( r ) n e ( 0 ) + T e e ln xe2x80x83 n e ( r ) n e ( 0 ) - ω e c B 0 r 2 2 - m e r 2 ω e 2 2 . ( 47 ) Using the definition of xcex2, ω e c B 0 β r 0 Δ xe2x80x83 r = ω e c 8 π ( n e0 T e + n i0 T i ) 2 ( T e + T i / 2 ) 1 2 4 π xe2x80x83 n e0 e 2 c ( ω i - ω e ) …
… = 4 ω e ω i - ω e ( n e0 T e + n i0 T i ) n e0 e . ( 48 ) Finally, using Eq. 48, the expressions for the electric potential and the flux function become Ψ ( r ) = - B 0 r 2 2 + c ω i - ω e ( n e0 T e + n i0 T i n e0 e ) ln xe2x80x83 n e ( r ) n e ( 0 ) xe2x80x83 and ( 49 ) Φ ( r ) = [ ω e ω i - ω e ( n e0 T e + n i0 T i ) n e0 e + T e e ] ln xe2x80x83 n e ( r ) n e ( 0 ) - ω e c B 0 r 2 2 - m e r 2 ω e 2 c . ( 50 ) Relationship Between xcfx89i and xcfx89e An expression for the electron angular velocity xcfx89i can also be derived from Eqs. 24 through 26. It is assumed that ions have an average energy xc2xdmi(rxcfx89i)2, which is determined by the method of formation of the FRC. Therefore, xcfx89i is determined by the FRC formation method, and xcfx89e can be determined by Eq. 24 by combining the equations for electrons and ions to eliminate the electric field: - [ n e m xe2x80x83 r xe2x80x83 ω e 2 + n i m i r xe2x80x83 ω i 2 ] = n e e xe2x80x83 r c ( ω i - ω e ) B z - T e ⅆ n e ⅆ r - T i ⅆ n i ⅆ r . ( 51 ) Eq. 25 can then be used to eliminate (xcfx89ixe2x88x92xcfx89e) to obtain [ n e m xe2x80x83 r xe2x80x83 ω e 2 + n i m i r xe2x80x83 ω i 2 ] = ⅆ ⅆ r ( B z 2 8 π + ∑ j n j T j ) . ( 52 ) Eq. 52 can be integrated from r=0 to rB={square root over (2)}r0. Assuming r0/xcex94r greater than greater than 1, the density is very small at both boundaries and Bz=xe2x88x92B0 (1xc2x1{square root over (xcex2)}). Carrying out the integration shows [ n e0 m xe2x80x83 ω e 2 + n i0 m i ω i 2 ] r 0 Δ xe2x80x83 r = B 0 2 π [ 8 π ( n e0 T e + n i0 T i ) ] 1 2 . ( 53 ) Using Eq. 33 for xcex94r yields an equation for xcfx89e: ω i 2 + Zm m i ω e 2 = Ω 0 ( ω i - ω e ) , where xe2x80x83 Ω 0 = ZeB 0 m i c . ( 54 ) Some limiting cases derived from Eq. 54 are: 1. xe2x80x83 ω i = 0 xe2x80x83 and xe2x80x83 ω e = - eB 0 mc ;xe2x80x832. xcfx89e=0 and xcfx89i=xcexa90; and 3. xe2x80x83 Z xe2x80x83 m m i ω e 2 ⪡ ω i 2 xe2x80x83 and xe2x80x83 ω e ≅ ω i ( 1 - ω i Ω 0 ) . In the first case, the current is carried entirely by electrons moving in their diamagnetic direction (xcfx89e less than 0). The electrons are confined magnetically, and the ions are confined electrostatically by E r = T i Zen i ⅆ n i ⅆ r xe2x80x83 ≤ 0 xe2x80x83 for xe2x80x83 r ≥ r 0 ≥ 0 xe2x80x83 for xe2x80x83 r ≤ r 0 . ( 55 ) In the second case, the current is carried entirely by ions moving in their diamagnetic direction (xcfx89i greater than 0). If xcfx89i is specified from the ion energy xc2xdmi(rxcfx891)2, determined in the formation process, then xcfx89e=0 and xcexa90=xcfx89e identifies the value of B0, the externally applied magnetic field. The ions are magnetically confined, and electrons are electrostatically confined by E r = T e en e ⅆ n e ⅆ r xe2x80x83 ≥ 0 xe2x80x83 for xe2x80x83 r ≥ r 0 ≤ 0 xe2x80x83 for xe2x80x83 r ≤ r 0 . ( 56 ) In the third case, xcfx89e greater than 0 and xcexa90 greater than xcfx89i. Electrons move in their counter diamagnetic direction and reduce the current density. From Eq. 33, the width of the distribution ni(r) is increased; however, the total current/unit length is I θ = ∫ r = 0 r B j θ ⅆ r = N e 2 π e ( ω i - ω e ) , where ( 57 ) N e = ∫ r = 0 r B 2 π xe2x80x83 r xe2x80x83 ⅆ rn e = 2 π xe2x80x83 r 0 Δ xe2x80x83 rn e0 . ( 58 ) xe2x80x83Ne=∫r=0rB2xcfx80rdrne=2xcfx80r0xcex94rne0.xe2x80x83xe2x80x83(58) Here, rB={square root over (2)}r0 and r0xcex94rxe2x89xa1(xcfx89ixe2x88x92xcfx89e)xe2x88x921 according to Eq. 33. The electron angular velocity xcfx89e can be increased by tuning the applied magnetic field B0. This does not change either Ixcex8 or the maximum magnetic field produced by the plasma current, which is B0{square root over (xcex2)}=(2xcfx80/c)Ixcex8. However, it does change xcex94r and, significantly, the potential "PHgr". The maximum value of "PHgr" is increased, as is the electric field that confines the electrons. Tuning the Magnetic Field In FIGS. 21A-D, the quantities ne/ne0 186, Bz/(B0{square root over (xcex2)}) 188, ("PHgr"/"PHgr"0 190, and xcexa8/xcexa80 192 are plotted against r/r0 194 for various values of B0. The values of potential and flux are normalized to ("PHgr"0=20(Te+Ti)/e and xcexa80=(c/xcfx89i)"PHgr"0. A deuterium plasma is assumed with the following data: ne0=ni0=1015 cmxe2x88x923; r0=40 cm; xc2xdmi(r0xcfx89i)2=300 keV; and Te=Ti=100 keV. For each of the cases illustrated in FIG. 21, xcfx89i=1.35xc3x97107 sxe2x88x921, and xcfx89e is determined from Eq. 54 for various values of B0: The case of xcfx89e=xe2x88x92xcfx89i and B0=1.385 kG involves magnetic confinement of both electrons and ions. The potential reduces to "PHgr"/"PHgr"0=mi(rxcfx89i)2/[80(Te+Ti)], which is negligible compared to the case xcfx893=0. The width of the density distribution xcex94r is reduced by a factor of 2, and the maximum magnetic field B0{square root over (xcex2)} is the same as for xcfx89e=0. Solution for Plasmas of Multiple Types of Ions This analysis can be carried out to include plasmas comprising multiple types of ions. Fusion fuels of interest involve two different kinds of ions, e.g., D-T, D-He3, and H-B11. The equilibrium equations (Eqs. 24 through 26) apply, except that j=e, 1, 2 denotes electrons and two types of ions where Z1=1 in each case and Z2=Z=1, 2, 5 for the above fuels. The equations for electrons and two types of ions cannot be solved exactly in terms of elementary functions. Accordingly, an iterative method has been developed that begins with an approximate solution. The ions are assumed to have the same values of temperature and mean velocity Vi=rxcfx89i. Ion-ion collisions drive the distributions toward this state, and the momentum transfer time for the ionxe2x80x94ion collisions is shorter than for ion-electron collisions by a factor of an order of 1000. By using an approximation, the problem with two types of ions can be reduced to a single ion problem. The momentum conservation equations for ions are - n 1 m 1 r xe2x80x83 ω 1 2 = n 1 e [ E r + r xe2x80x83 ω 1 c B z ] - T i ⅆ n i ⅆ r xe2x80x83 and ( 59 ) - n 2 m 2 r xe2x80x83 ω 2 2 = n 2 Ze [ E r + r xe2x80x83 ω 2 c B z ] - T 2 ⅆ n 2 ⅆ r . ( 60 ) In the present case, T1=T2 and xcfx891=xcfx892. Adding these two equations results in - n 1 ⟨ m i ⟩ ω i 2 = n i ⟨ Z ⟩ e [ E r + r xe2x80x83 ω i c B z ] - T i ⅆ n i ⅆ r , ( 61 ) where ni=n1=n2; xcfx89i=xcfx891=xcfx892; Ti=T1=T2; ni(mi)=n1m1+n2m2; and ni(Z)=n1+n2Z. The approximation is to assume that (mi) and (z) are constants obtained by replacing ni(r) and n2(r) by n10 and n20, the maximum values of the respective functions. The solution of this problem is now the same as the previous solution for the single ion type, except that (Z) replaces Z and (mi) replaces mi. The values of n1 and n2 can be obtained from n1+n2=ni and n1+Zn2=ne=(Z)ni. It can be appreciated that n1 and n2 have the same functional form. Now the correct solution can be obtained by iterating the equations: ⅆ log xe2x80x83 N 1 ⅆ ξ = m 1 r 0 2 Ω 1 ( ω i - ω e ) T i B z ( ξ ) B 0 - T e T i xe2x80x83 ⅆ log xe2x80x83 N e ⅆ ξ + m 1 ( ω i r 0 ) 2 T i xe2x80x83 and ( 62 ) ⅆ log xe2x80x83 N 2 ⅆ ξ = m 2 r 0 2 Ω 2 ( ω i - ω e ) T i B z ( ξ ) B 0 - Z xe2x80x83 T e T i xe2x80x83 ⅆ log xe2x80x83 N e ⅆ ξ + m 2 ( ω i r 0 ) 2 T i ,
w xe2x80x83 h xe2x80x83 e xe2x80x83 r xe2x80x83 e
N 1 = n 1 ( r ) n 10 , N 2 = n 2 ( r ) n 20 , ξ = r 2 2 r 0 2 , Ω 1 = e xe2x80x83 B 0 m 1 c , a xe2x80x83 n xe2x80x83 d xe2x80x83 Ω 2 = Z xe2x80x83 e xe2x80x83 B 0 m 2 c . ( 63 ) The first iteration can be obtained by substituting the approximate values of Bz("xgr") and Ne("xgr") in the right hand sides of Eqs. 62 and 63 and integrating to obtain the corrected values of n1(r), n2(r), and Bz(r). Calculations have been carried out for the data shown in Table 1, below. Numerical results for fusion fuels are shown in FIGS. 22A-D through 24A-D wherein the quantities n1/n10 206, "PHgr"/"PHgr"0 208, and xcexa8/xcexa80 210 are plotted against r/r0 204. FIGS. 22A-D shows the first approximation (solid lines) and the final results (dotted lines) of the iterations for D-T for the normalized density of D 196, the normalized density of T 198, the normalized electric potential 200, and the normalized flux 202. FIGS. 23A-D show the same iterations for D-He3 for the normalized density of D 212, the normalized density of He3 214, the normalized electric potential 216, and the normalized flux 218. FIGS. 24A-D show the same iterations for p-B11 for the normalized density of p 220, the normalized density of B11 222, the normalized electric potential 224, and the normalized flux 226. Convergence of the iteration is most rapid for D-T. In all cases the first approximation is close to the final result. Structure of the Containment System FIG. 25 illustrates a preferred embodiment of a containment system 300 according to the present invention. The containment system 300 comprises a chamber wall 305 that defines therein a confining chamber 310. Preferably, the chamber 310 is cylindrical in shape, with principle axis 315 along the center of the chamber 310. For application of this containment system 300 to a fusion reactor, it is necessary to create a vacuum or near vacuum inside the chamber 310. Concentric with the principle axis 315 is a betatron flux coil 320, located within the chamber 310. The betatron flux coil 320 comprises an electrical current carrying medium adapted to direct current around a long coil, as shown, which preferably comprises parallel winding multiple separate coils, and most perferably parallel windings of about four separate coils, to form a long coil. Persons skilled in the art will appreciate that current through the betatron coil 320 will result in a magnetic field inside the betatron coil 320, substantially in the direction of the principle axis 315. Around the outside of the chamber wall 305 is an outer coil 325. The outer coil 325 produce a relatively constant magnetic field having flux substantially parallel with principle axis 315. This magnetic field is azimuthally symmetrical. The approximation that the magnetic field due to the outer coil 325 is constant and parallel to axis 315 is most valid away from the ends of the chamber 310. At each end of the chamber 310 is a mirror coil 330. The mirror coils 330 are adapted to produce an increased magnetic field inside the chamber 310 at each end, thus bending the magnetic field lines inward at each end. (See FIGS. 8 and 10.) As explained, this bending inward of the field lines helps to contain the plasma 335 in a containment region within the chamber 310 generally between the mirror coils 330 by pushing it away from the ends where it can escape the containment system 300. The mirror coils 330 can be adapted to produce an increased magnetic field at the ends by a variety of methods known in the art, including increasing the number of windings in the mirror coils 330, increasing the current through the mirror coils 330, or overlapping the mirror coils 330 with the outer coil 325. The outer coil 325 and mirror coils 330 are shown in FIG. 25 implemented outside the chamber wall 305; however, they may be inside the chamber 310. In cases where the chamber wall 305 is constructed of a conductive material such as metal, it may be advantageous to place the coils 325, 330 inside the chamber wall 305 because the time that it takes for the magnetic field to diffuse through the wall 305 may be relatively large and thus cause the system 300 to react sluggishly. Similarly, the chamber 310 may be of the shape of a hollow cylinder, the chamber wall 305 forming a long, annular ring. In such a case, the betatron flux coil 320 could be implemented outside of the chamber wall 305 in the center of that annular ring. Preferably, the inner wall forming the center of the annular ring may comprise a non-conducting material such as glass. As will become apparent, the chamber 310 must be of sufficient size and shape to allow the circulating plasma beam or layer 335 to rotate around the principle axis 315 at a given radius. The chamber wall 305 may be formed of a material having a high magnetic permeability, such as steel. In such a case, the chamber wall 305, due to induced countercurrents in the material, helps to keep the magnetic flux from escaping the chamber 310, xe2x80x9ccompressingxe2x80x9d it. If the chamber wall were to be made of a material having low magnetic permeability, such as plexiglass, another device for containing the magnetic flux would be necessary. In such a case, a series of closed-loop, flat metal rings could be provided. These rings, known in the art as flux delimiters, would be provided within the outer coils 325 but outside the circulating plasma beam 335. Further, these flux delimiters could be passive or active, wherein the active flux delimiters would be driven with a predetermined current to greater facilitate the containment of magnetic flux within the chamber 310. Alternatively, the outer coils 325 themselves could serve as flux delimiters. As explained above, a circulating plasma beam 335, comprising charged particles, may be contained within the chamber 310 by the Lorentz force caused by the magnetic field due to the outer coil 325. As such, the ions in the plasma beam 335 are magnetically contained in large betatron orbits about the flux lines from the outer coil 325, which are parallel to the principle axis 315. One or more beam injection ports 340 are also provided for adding plasma ions to the circulating plasma beam 335 in the chamber 310. In a preferred embodiment, the injector ports 340 are adapted to inject an ion beam at about the same radial position from the principle axis 315 where the circulating plasma beam 335 is contained (i.e., around the null surface). Further, the injector ports 340 are adapted to inject ion beams 350 (See FIG. 28) tangent to and in the direction of the betatron orbit of the contained plasma beam 335. Also provided are one or more background plasma sources 345 for injecting a cloud of non-energetic plasma into the chamber 310. In a preferred embodiment, the background plasma sources 345 are adapted to direct plasma 335 toward the axial center of the chamber 310. It has been found that directing the plasma this way helps to better contain the plasma 335 and leads to a higher density of plasma 335 in the containment region within the chamber 310. Formation of the FRC Conventional procedures used to form a FRC primarily employ the theta pinch-field reversal procedure. In this conventional method, a bias magnetic field is applied by external coils surrounding a neutral gas back-filled chamber. Once this has occurred, the gas is ionized and the bias magnetic field is frozen in the plasma. Next, the current in the external coils is rapidly reversed and the oppositely oriented magnetic field lines connect with the previously frozen lines to form the closed topology of the FRC (see FIG. 8). This formation process is largely empirical and there exists almost no means of controlling the formation of the FRC. The method has poor reproducibility and no tuning capability as a result. In contrast, the FRC formation methods of the present invention allow for ample control and provide a much more transparent and reproducible process. In fact, the FRC formed by the methods of the present invention can be tuned and its shape as well as other properties can be directly influenced by manipulation of the magnetic field applied by the outer field coils 325. Formation of the FRC by methods of the present inventions also results in the formation of the electric field and potential well in the manner described in detail above. Moreover, the present methods can be easily extended to accelerate the FRC to reactor level parameters and high-energy fuel currents, and advantageously enables the classical confinement of the ions. Furthermore, the technique can be employed in a compact device and is very robust as well as easy to implementxe2x80x94all highly desirable characteristics for reactor systems. In the present methods, FRC formation relates to the circulating plasma beam 335. It can be appreciated that the circulating plasma beam 335, because it is a current, creates a poloidal magnetic field, as would an electrical current in a circular wire. Inside the circulating plasma beam 335, the magnetic self-field that it induces opposes the externally applied magnetic field due to the outer coil 325. Outside the plasma beam 335, the magnetic self-field is in the same direction as the applied magnetic field. When the plasma ion current is sufficiently large, the self-field overcomes the applied field, and the magnetic field reverses inside the circulating plasma beam 335, thereby forming the FRC topology as shown in FIGS. 8 and 10. The requirements for field reversal can be estimated with a simple model. Consider an electric current Ip carried by a ring of major radius r0 and minor radius a less than less than r0. The magnetic field at the center of the ring normal to the ring is Bp=2xcfx80Ip/(cr0). Assume that the ring current Ip=Npe(xcexa90/2xcfx80) is carried by Np ions that have an angular velocity xcexa90. For a single ion circulating at radius r0=V0/xcexa90,xcexa90=eB0/mic is the cyclotron frequency for an external magnetic field B0. Assume V0 is the average velocity of the beam ions. Field reversal is defined as B p = N p e xe2x80x83 Ω 0 r 0 c ≥ 2 B 0 , ( 64 ) which implies that Np greater than 2 r0/ai, and I p ≥ e xe2x80x83 V 0 πα i , ( 65 ) where xcex11=e2/mic2=1.57xc3x9710xe2x88x9216 cm and the ion beam energy is 1 2 m i V 0 2 . In the one dimensional model, the magnetic field from the plasma current is Bp=(2xcfx80/c)ip, where ip is current per unit of length. The field reversal requirement is ip greater than eV0/xcfx80r0xcex1i=0.225 kA/cm, where B0=69.3 G and 1 2 m i V 0 2 = 100 xe2x80x83 eV . For a model with periodic rings and Bz is averaged over the axial coordinate (Bz)=(2xcfx80/c)(Ip/s) (s is the ring spacing), if s=r0, this model would have the same average magnetic field as the one dimensional model with ip=Ip/s. Combined Beam/Betatron Formation Technique A preferred method of forming a FRC within the confinement system 300 described above is herein termed the combined beam/betatron technique. This approach combines low energy beams of plasma ions with betatron acceleration using the betatron flux coil 320. The first step in this method is to inject a substantially annular cloud layer of background plasma in the chamber 310 using the background plasma sources 345. Outer coil 325 produces a magnetic field inside the chamber 310, which magnetizes the background plasma. At short intervals, low energy ion beams are injected into the chamber 310 through the injector ports 340 substantially transverse to the externally applied magnetic field within the chamber 310. As explained above, the ion beams are trapped within the chamber 310 in large betatron orbits by this magnetic field. The ion beams may be generated by an ion accelerator, such as an accelerator comprising an ion diode and a Marx generator. (see R. B. Miller, An Introduction to the Physics of Intense Charged Particle Beams, (1982)). As one of skill in the art can appreciate, the externally applied magnetic field will exert a Lorentz force on the injected ion beam as soon as it enters the chamber 310; however, it is desired that the beam not deflect, and thus not enter a betatron orbit, until the ion beam reaches the circulating plasma beam 335. To solve this problem, the ion beams are neutralized with electrons and directed through a substantially constant unidirectional magnetic field before entering the chamber 310. As illustrated in FIG. 26, when the ion beam 350 is directed through an appropriate magnetic field, the positively charged ions and negatively charged electrons separate. The ion beam 350 thus acquires an electric self-polarization due to the magnetic field. This magnetic field may be produced by, e.g., a permanent magnet or by an electromagnet along the path of the ion beam. When subsequently introduced into the confinement chamber 310, the resultant electric field balances the magnetic force on the beam particles, allowing the ion beam to drift undeflected. FIG. 27 shows a head-on view of the ion beam 350 as it contacts the plasma 335. As depicted, electrons from the plasma 335 travel along magnetic field lines into or out of the beam 350, which thereby drains the beam""s electric polarization. When the beam is no longer electrically polarized, the beam joins the circulating plasma beam 335 in a betatron orbit around the principle axis 315, as shown in FIG. 25. When the plasma beam 335 travels in its betatron orbit, the moving ions comprise a current, which in turn gives rise to a poloidal magnetic self-field. To produce the FRC topology within the chamber 310, it is necessary to increase the velocity of the plasma beam 335, thus increasing the magnitude of the magnetic self-field that the plasma beam 335 causes. When the magnetic self-field is large enough, the direction of the magnetic field at radial distances from the axis 315 within the plasma beam 335 reverses, giving rise to a FRC. (See FIGS. 8 and 10). It can be appreciated that, to maintain the radial distance of the circulating plasma beam 335 in the betatron orbit, it is necessary to increase the applied magnetic field from the outer coil 325 as the plasma beam 335 increases in velocity. A control system is thus provided for maintaining an appropriate applied magnetic field, dictated by the current through the outer coil 325. Alternatively, a second outer coil may be used to provide the additional applied magnetic field that is required to maintain the radius of the plasma beam""s orbit as it is accelerated. To increase the velocity of the circulating plasma beam 335 in its orbit, the betatron flux coil 320 is provided. Referring to FIG. 28, it can be appreciated that increasing a current through the betatron flux coil 320, by Ampere""s Law, induces an azimuthal electric field, E, inside the chamber 310. The positively charged ions in the plasma beam 335 are accelerated by this induced electric field, leading to field reversal as described above. When ion beams are added to the circulating plasma beam 335, as described above, the plasma beam 335 depolarizes the ion beams. For field reversal, the circulating plasma beam 335 is preferably accelerated to a rotational energy of about 100 eV, and preferably in a range of about 75 eV to 125 eV. To reach fusion relevant conditions, the circulating plasma beam 335 is preferably accelerated to about 200 keV and preferably to a range of about 100 keV to 3.3 MeV. In developing the necessary expressions for the betatron acceleration, the acceleration of single particles is first considered. The gyroradius of ions r=V/xcexa9i will change because V increases and the applied magnetic field must change to maintain the radius of the plasma beam""s orbit, r0=V/xcexa9c ∂ r ∂ t = 1 Ω [ ∂ V ∂ t - V Ω i ∂ Ω i ∂ t ] = 0 ,
w xe2x80x83 h xe2x80x83 e xe2x80x83 r xe2x80x83 e ( 66 ) ∂ V ∂ t = r 0 e m i c ∂ B c ∂ t = e xe2x80x83 E θ m i = - e m i c 1 2 π xe2x80x83 r 0 ∂ Ψ ∂ t , ( 67 ) and xcexa8 is the magnetic flux: Ψ = ∫ 0 r 0 B z 2 xe2x80x83 π xe2x80x83 r ⅆ r = π xe2x80x83 r 0 2 ⟨ B z ⟩ , ( 68 ) w xe2x80x83 h xe2x80x83 e xe2x80x83 r xe2x80x83 e xe2x80x83 ⟨ B z ⟩ = - B F ( r a r 0 ) 2 - B c [ 1 - ( r a r 0 ) 2 ] . ( 69 ) From Eq. 67, it follows that ∂ ⟨ B z ⟩ ∂ t = - 2 ∂ B c ∂ t , ( 70 ) and (Bz)=xe2x88x922Bc+B0, assuming that the initial values of BF and Bc are both B0. Eq. 67 can be expressed as ∂ V ∂ t = - e 2 m i c r 0 ∂ ⟨ B z ⟩ ∂ t . ( 71 ) After integration from the initial to final states where 1 2 m xe2x80x83 V 0 2 = W 0 xe2x80x83 a xe2x80x83 n xe2x80x83 d xe2x80x83 1 2 m xe2x80x83 V 2 = W , the final values of the magnetic fields are: B c = B 0 W W 0 = 2.19 xe2x80x83 kG
a xe2x80x83 n xe2x80x83 d ( 72 ) B F = B 0 [ W W 0 + ( r 0 r a ) 2 ( W W 0 - 1 ) ] = 10.7 xe2x80x83 kG , ( 73 ) assuming B0=69.3 G, W/W0=1000, and r0/ra=2. This calculation applies to a collection of ions, provided that they are all located at nearly the same radius r0 and the number of ions is insufficient to alter the magnetic fields. The modifications of the basic betatron equations to accommodate the present problem will be based on a one-dimensional equilibrium to describe the multi-ring plasma beam, assuming the rings have spread out along the field lines and the z-dependence can be neglected. The equilibrium is a self-consistent solution of the Vlasov-Maxwell equations that can be summarized as follows: (a) The density distribution is n = n m cosh 2 ( r 2 - r 0 2 r 0 Δ xe2x80x83 r ) , ( 74 ) which applies to the electrons and protons (assuming quasi neutrality); r0 is the position of the density maximum; and xcex94r is the width of the distribution; and (b) The magnetic field is B z = - B c - 2 π xe2x80x83 I p c tanh ( r 2 - r 0 2 r 0 Δ xe2x80x83 r ) , ( 75 ) where Bc is the external field produced by the outer coil 325. Initially, Bc=B0. This solution satisfies the boundary conditions that r=ra, and r=rb are conductors (Bnormal=0) and equipotentials with potential "PHgr"=0. The boundary conditions are satisfied if r02=(ra2+rb2)/2. ra=10 cm and r0=20 cm, so it follows that rb=26.5 cm. Ip is the plasma current per unit length. The average velocities of the beam particles are Vi=r0xcfx89i and Ve=r0xcfx89e, which are related by the equilibrium condition: ω e = ω i ( 1 - ω i Ω i ) , ( 76 ) where xcexa9i=eBc/(mic). Initially, it is assumed Bc=B0, xcfx89i=xcexa9iand xcfx89e=0. (In the initial equilibrium there is an electric field such that the {right arrow over (E)}xc3x97{right arrow over (B)} and the xcex94Bxc3x97{right arrow over (B)} drifts cancel. Other equilibrium are possible according to the choice of Bc.) The equilibrium equations are assumed to be valid if xcfx89i and Bc are slowly varying functions of time, but r0=Vi/xcexa9i remains constant. The condition for this is the same as Eq. 66. Eq. 67 is also similar, but the flux function xcexa8 has an additional term, i.e., xcexa8=xcfx80r02(Bz) where ⟨ B z ⟩ = B _ z + 2 π c I p ( r b 2 - r a 2 r b 2 + r a 2 )
a xe2x80x83 n xe2x80x83 d ( 77 ) B _ z = - B F ( r a r 0 ) 2 - B c [ 1 - ( r a r 0 ) 2 ] . ( 78 ) The magnetic energy per unit length due to the beam current is ∫ r a r b 2 π xe2x80x83 r ⅆ r ( B z - B c 8 π ) 2 = 1 2 L p I p 2 , ( 79 ) f xe2x80x83 r xe2x80x83 o xe2x80x83 m xe2x80x83 w xe2x80x83 h xe2x80x83 i xe2x80x83 c xe2x80x83 h xe2x80x83 L p = r b 2 - r a 2 r b 2 + r a 2 2 π 2 r 0 2 c 2 xe2x80x83 a xe2x80x83 n xe2x80x83 d xe2x80x83 ⟨ B z ⟩ = B _ z + c π xe2x80x83 r 0 2 L p I p . ( 80 ) The betatron condition of Eq. 70 is thus modified so that ∂ B _ z ∂ t = - 2 ∂ B c ∂ t - L p c π xe2x80x83 r 0 2 ∂ I p ∂ t , ( 81 ) and xe2x80x83 Eq . xe2x80x83 67 xe2x80x83 becomes : xe2x80x83 ∂ V i ∂ t = e m i r 0 c ∂ B c ∂ t = - e 2 m i c r 0 ∂ B _ z ∂ t - e m i L p 2 π xe2x80x83 r 0 ∂ I p ∂ t . ( 82 ) After xe2x80x83 integrating , xe2x80x83 Δ B _ z = - 2 B 0 [ 1 + r b 2 - r a 2 r 0 2 ] [ W W 0 - 1 ] . ( 83 ) For W0=100 eV and W=100 keV, xcex94{right arrow over (B)}z=xe2x88x927.49 kG. Integration of Eqs. 81 and 82 determines the value of the magnetic field produced by the field coil: B c = B 0 W W 0 = 2.19 xe2x80x83 kG ( 84 ) and xe2x80x83 B F = B F0 - ( r 0 r a ) 2 Δ xe2x80x83 B _ z - ( r 0 2 - r a 2 r a 2 ) Δ xe2x80x83 B c = 25 xe2x80x83 kG . ( 85 ) If the final energy is 200 keV, Bc=3.13 kG and BF=34.5 kG. The magnetic energy in the flux coil would be B F 2 8 π π xe2x80x83 r F 2 l = 172 xe2x80x83 k xe2x80x83 J . The plasma current is initially 0.225 kA/cm corresponding to a magnetic field of 140 G, which increases to 10 kA/cm and a magnetic field of 6.26 kG. In the above calculations, the drag due to Coulomb collisions has been neglected. In the injection/trapping phase, it was equivalent to 0.38 volts/cm. It decreases as the electron temperature increases during acceleration. The inductive drag, which is included, is 4.7 volts/cm, assuming acceleration to 200 keV in 100 xcexcs. The betatron flux coil 320 also balances the drag from collisions and inductance. The frictional and inductive drag can be described by the equation: ∂ V b ∂ t = - V b [ 1 t be + 1 t bi ] - e m b L 2 xe2x80x83 r 0 ∂ I b ∂ t , ( 86 ) where (Ti/mi) less than Vb less than (Tc/m). Here, Vb is the beam velocity, Te and Ti are electron and ion temperatures, Ib is the beam ion current, and L = 0.01257 r 0 [ ln ( 8 r 0 a ) - 7 4 ] = 0.71 µ xe2x80x83 H is the ring inductance. Also, r0=20 cm and a=4 cm. The Coulomb drag is determined by t be = 3 4 2 π ( m i m ) T e 3 / 2 ne 4 ln xe2x80x83 Λ = 195 xe2x80x83 µ xe2x80x83 sec
t bi = 2 2 m i W b 3 / 2 4 π xe2x80x83 ne 4 ln xe2x80x83 Λ = 54.8 xe2x80x83 µ xe2x80x83 sec ( 87 ) To compensate the drag, the betatron flux coil 320 must provide an electric field of 1.9 volts/cm (0.38 volts/cm for the Coulomb drag and 1.56 volts/cm for the inductive drag). The magnetic field in the betatron flux coil 320 must increase by 78 Gauss/xcexcs to accomplish this, in which case Vb will be constant. The rise time of the current to 4.5 kA is 18 xcexcs, so that the magnetic field BF will increase by 1.4 kG. The magnetic field energy required in the betatron flux coil 320 is B F 2 8 π xc3x97 xe2x80x83 r F 2 l = 394 xe2x80x83 Joules xe2x80x83 ( l = 115 xe2x80x83 cm ) . ( 88 ) Betatron Formation Technique Another preferred method of forming a FRC within the confinement system 300 is herein termed the betatron formation technique. This technique is based on driving the betatron induced current directly to accelerate a circulating plasma beam 335 using the betatron flux coil 320. A preferred embodiment of this technique uses the confinement system 300 depicted in FIG. 25, except that the injection of low energy ion beams is not necessary. As indicated, the main component in the betatron formation technique is the betatron flux coil 320 mounted in the center and along the axis of the chamber 310. Due to its separate parallel windings construction, the coil 320 exhibits very low inductance and, when coupled to an adequate power source, has a low LC time constant, which enables rapid ramp up of the current in the flux coil 320. Preferably, formation of the FRC commences by energizing the external field coils 325, 330. This provides an axial guide field as well as radial magnetic field components near the ends to axially confine the plasma injected into the chamber 310. Once sufficient magnetic field is established, the background plasma sources 345 are energized from their own power supplies. Plasma emanating from the guns streams along the axial guide field and spreads slightly due to its temperature. As the plasma reaches the mid-plane of the chamber 310, a continuous, axially extending, annular layer of cold, slowly moving plasma is established. At this point the betatron flux coil 320 is energized. The rapidly rising current in the coil 320 causes a fast changing axial flux in the coil""s interior. By virtue of inductive effects this rapid increase in axial flux causes the generation of an azimuthal electric field E (see FIG. 29), which permeates the space around the flux coil. By Maxwell""s equations, this electric field is directly proportional to the change in strength of the magnetic flux inside the coil, i.e.: a faster betatron coil current ramp-up will lead to a stronger electric field. The inductively created electric field couples to the charged particles in the plasma and causes a ponderomotive force, which accelerates the particles in the annular plasma layer. Electrons, by virtue of their smaller mass, are the first species to experience acceleration. The initial current formed by this process is, thus, primarily due to electrons. However, sufficient acceleration time (around hundreds of micro-seconds) will eventually also lead to ion current. Referring to FIG. 29, this electric field accelerates the electrons and ions in opposite directions. Once both species reach their terminal velocities, current is carried about equally by ions and electrons. As noted above, the current carried by the rotating plasma gives rise to a self magnetic field. The creation of the actual FRC topology sets in when the self magnetic field created by the current in the plasma layer becomes comparable to the applied magnetic field from the external field coils 325, 330. At this point magnetic reconnection occurs and the open field lines of the initial externally produced magnetic field begin to close and form the FRC flux surfaces (see FIGS. 8 and 10). The base FRC established by this method exhibits modest magnetic field and particle energies that are typically not at reactor relevant operating parameters. However, the inductive electric acceleration field will persist, as long as the current in the betatron flux coil 320 continues to increase at a rapid rate. The effect of this process is that the energy and total magnetic field strength of the FRC continues to grow. The extent of this process is, thus, primarily limited by the flux coil power supply, as continued delivery of current requires a massive energy storage bank. However, it is, in principal, straightforward to accelerate the system to reactor relevant conditions. For field reversal, the circulating plasma beam 335 is preferably accelerated to a rotational energy of about 100 eV, and preferably in a range of about 75 eV to 125 eV. To reach fusion relevant conditions, the circulating plasma beam 335 is preferably accelerated to about 200 keV and preferably to a range of about 100 keV to 3.3 MeV. When ion beams are added to the circulating plasma beam 335, as described above, the plasma beam 335 depolarizes the ion beams. Experimentsxe2x80x94Beam Trapping and FRC Formation Experiment 1: Propagating and trapping of a neutralized beam in a magnetic containment vessel to create an FRC. Beam propagation and trapping were successfully demonstrated at the following parameter levels: Vacuum chamber dimensions: about 1 m diameter, 1.5 m length. Betatron coil radius of 10 cm. Plasma beam orbit radius of 20 cm. Mean kinetic energy of streaming beam plasma was measured to be about 100 eV, with a density of about 103 cmxe2x88x923, kinetic temperature on the order of 10 eV and a pulse-length of about 20 xcexcs. Mean magnetic field produced in the trapping volume was around 100 Gauss, with a ramp-up period of 150 xcexcs. Source: Outer coils and betatron coils. Neutralizing background plasma (substantially Hydrogen gas) was characterized by a mean density of about 1013 cmxe2x88x923, kinetic temperature of less than 10 eV. The beam was generated in a deflagration type plasma gun. The plasma beam source was neutral Hydrogen gas, which was injected through the back of the gun through a special puff valve. Different geometrical designs of the electrode assembly were utilized in an overall cylindrical arrangement. The charging voltage was typically adjusted between 5 and 7.5 kV. Peak breakdown currents in the guns exceeded 250,000 A. During part of the experimental runs, additional pre-ionized plasma was provided by means of an array of small peripheral cable guns feeding into the central gun electrode assembly before, during or after neutral gas injection. This provided for extended pulse lengths of above 25 xcexcs. The emerging low energy neutralized beam was cooled by means of streaming through a drift tube of non-conducting material before entering the main vacuum chamber. The beam plasma was also pre-magnetized while streaming through this tube by means of permanent magnets. The beam self-polarized while traveling through the drift tube and entering the chamber, causing the generation of a beam-internal electric field that offset the magnetic field forces on the beam. By virtue of this mechanism it was possible to propagate beams as characterized above through a region of magnetic field without deflection. Upon further penetration into the chamber, the beam reached the desired orbit location and encountered a layer of background plasma provided by an array of cable guns and other surface flashover sources. The proximity of sufficient electron density caused the beam to loose its self-polarization field and follow single particle like orbits, essentially trapping the beam. Faraday cup and B-dot probe measurements confirmed the trapping of the beam and its orbit. The beam was observed to have performed the desired circular orbit upon trapping. The beam plasma was followed along its orbit for close to xc2xe of a turn. The measurements indicated that continued frictional and inductive losses caused the beam particles to loose sufficient energy for them to curl inward from the desired orbit and hit the betatron coil surface at around the {fraction (3/4 )} turn mark. To prevent this, the losses could be compensated by supplying additional energy to the orbiting beam by inductively driving the particles by means of the betatron coil. Experiment 2: FRC formation utilizing the combined beam/betatron formation technique. FRC formation was successfully demonstrated utilizing the combined beam/betatron formation technique. The combined beam/betatron formation technique was performed experimentally in a chamber 1 m in diameter and 1.5 m in length using an externally applied magnetic field of up to 500 G, a magnetic field from the betatron flux coil 320 of up to 5 kG, and a vacuum of 1.2xc3x9710xe2x88x925 torr. In the experiment, the background plasma had a density of 1013 cmxe2x88x923 and the ion beam was a neutralized Hydrogen beam having a density of 1.2xc3x971013 cmxe2x88x923, a velocity of 2xc3x97107 cm/s, and a pulse length of around 20 xcexcs (at half height). Field reversal was observed. Experiment 3: FRC formation utilizing the betatron formation technique. FRC formation utilizing the betatron formation technique was successfully demonstrated at the following parameter levels: Vacuum chamber dimensions: about 1 m diameter, 1.5 m length. Betatron coil radius of 10 cm. Plasma-orbit radius of 20 cm. Mean external magnetic field produced in the vacuum chamber was up to 100 Gauss, with a ramp-up period of 150 xcexcs and a mirror ratio of 2 to 1. (Source: Outer coils and betatron coils). The background plasma (substantially Hydrogen gas) was characterized by a mean density of about 1013 cmxe2x88x923, kinetic temperature of less than 10 eV. The lifetime of the configuration was limited by the total energy stored in the experiment and generally was around 30 xcexcs. The experiments proceeded by first injecting a background plasma layer by two sets of coaxial cable guns mounted in a circular fashion inside the chamber. Each collection of 8 guns was mounted on one of the two mirror coil assemblies. The guns were azimuthally spaced in an equidistant fashion and offset relative to the other set. This arrangement allowed for the guns to be fired simultaneously and thereby created an annular plasma layer. Upon establishment of this layer, the betatron flux coil was energized. Rising current in the betatron coil windings caused an increase in flux inside the coil, which gave rise to an azimuthal electric field curling around the betatron coil. Quick ramp-up and high current in the betatron flux coil produced a strong electric field, which accelerated the annular plasma layer and thereby induced a sizeable current. Sufficiently strong plasma current produced a magnetic self-field that altered the externally supplied field and caused the creation of the field reversed configuration. Detailed measurements with B-dot loops identified the extent, strength and duration of the FRC. An example of typical data is shown by the traces of B-dot probe signals in FIG. 30. The data curve A represents the absolute strength of the axial component of the magnetic field at the axial mid-plane (75 cm from either end plate) of the experimental chamber and at a radial position of 15 cm. The data curve B represents the absolute strength of the axial component of the magnetic field at the chamber axial mid-plane and at a radial position of 30 cm. The curve A data set, therefore, indicates magnetic field strength inside of the fuel plasma layer (between betatron coil and plasma) while the curve B data set depicts the magnetic field strength outside of the fuel plasma layer. The data clearly indicates that the inner magnetic field reverses orientation (is negative) between about 23 and 47 xcexcs, while the outer field stays positive, i.e., does not reverse orientation. The time of reversal is limited by the ramp-up of current in the betatron coil. Once peak current is reached in the betatron coil, the induced current in the fuel plasma layer starts to decrease and the FRC rapidly decays. Up to now the lifetime of the FRC is limited by the energy that can be stored in the experiment. As with the injection and trapping experiments, the system can be upgraded to provide longer FRC lifetime and acceleration to reactor relevant parameters. Overall, this technique not only produces a compact FRC, but it is also robust and straightforward to implement. Most importantly, the base FRC created by this method can be easily accelerated to any desired level of rotational energy and magnetic field strength. This is crucial for fusion applications and classical confinement of high-energy fuel beams. Experiment 4: FRC formation utilizing the betatron formation technique. An attempt to form an FRC utilizing the betatron formation technique has been performed experimentally in a chamber 1 m in diameter and 1.5 m in length using an externally applied magnetic field of up to 500 G, a magnetic field from the betatron flux coil 320 of up to 5 kG, and a vacuum of 5xc3x9710xe2x88x926 torr. In the experiment, the background plasma comprised substantially Hydrogen with of a density of 1013 cmxe2x88x923 and a lifetime of about 40 xcexcs. Field reversal was observed. Fusion Significantly, these two techniques for forming a FRC inside of a containment system 300 described above, or the like, can result in plasmas having properties suitable for causing nuclear fusion therein. More particularly, the FRC formed by these methods can be accelerated to any desired level of rotational energy and magnetic field strength. This is crucial for fusion applications and classical confinement of high-energy fuel beams. In the confinement system 300, therefore, it becomes possible to trap and confine high-energy plasma beams for sufficient periods of time to cause a fusion reaction therewith. To accommodate fusion, the FRC formed by these methods is preferably accelerated to appropriate levels of rotational energy and magnetic field strength by betatron acceleration. Fusion, however, tends to require a particular set of physical conditions for any reaction to take place. In addition, to achieve efficient bum-up of the fuel and obtain a positive energy balance, the fuel has to be kept in this state substantially unchanged for prolonged periods of time. This is important, as high kinetic temperature and/or energy characterize a fusion relevant state. Creation of this state, therefore, requires sizeable input of energy, which can only be recovered if most of the fuel undergoes fusion. As a consequence, the confinement time of the fuel has to be longer than its bum time. This leads to a positive energy balance and consequently net energy output. A significant advantage of the present invention is that the confinement system and plasma described herein are capable of long confinement times, i.e., confinement times that exceed fuel bum times. A typical state for fusion is, thus, characterized by the following physical conditions (which tend to vary based on fuel and operating mode): Average ion temperature: in a range of about 30 to 230 keV and preferably in a range of about 80 keV to 230 keV Average electron temperature: in a range of about 30 to 100 keV and preferably in a range of about 80 to 100 keV Coherent energy of the fuel beams (injected ion beams and circulating plasma beam): in a range of about 100 keV to 3.3 MeV and preferably in a range of about 300 keV to 3.3 MeV. Total magnetic field: in a range of about 47.5 to 120 kG and preferably in a range of about 95 to 120 kG (with the externally applied field in a range of about 2.5 to 15 kG and preferably in a range of about 5 to 15 kG). Classical Confinement time: greater than the fuel burn time and preferably in a range of about 10 to 100 seconds. Fuel ion density: in a range of about 1014 to less than 1016 cmxe2x88x923 and preferably in a range of about 1014 to 1015 cm3. Total Fusion Power: preferably in a range of about 50 to 450 kW/cm (power per cm of chamber length) To accommodate the fusion state illustrated above, the FRC is preferably accelerated to a level of coherent rotational energy preferably in a range of about 100 keV to 3.3 MeV, and more preferably in a range of about 300 keV to 3.3 MeV, and a level of magnetic field strength preferably in a range of about 45 to 120 kG, and more preferably in a range of about 90 to 115 kG. At these levels, high energy ion beams can be injected into the FRC and trapped to form a plasma beam layer wherein the plasma beam ions are magnetically confined and the plasma beam electrons are electrostatically confined. Preferably, the electron temperature is kept as low as practically possible to reduce the amount of bremsstrahlung radiation, which can, otherwise, lead to radiative energy losses. The electrostatic energy well of the present invention provides an effective means of accomplishing this. The ion temperature is preferably kept at a level that provides for efficient burn-up since the fusion cross-section is a function of ion temperature. High direct energy of the fuel ion beams is essential to provide classical transport as discussed in this application. It also minimizes the effects of instabilities on the fuel plasma. The magnetic field is consistent with the beam rotation energy. It is partially created by the plasma beam (self-field) and in turn provides the support and force to keep the plasma beam on the desired orbit. Fusion Products The fusion products are born predominantly near the null surface from where they emerge by diffusion towards the separatrix 84 (see FIG. 8). This is due to collisions with electrons (as collisions with ions do not change the center of mass and therefore do not cause them to change field lines). Because of their high kinetic energy (product ions have much higher energy than the fuel ions), the fusion products can readily cross the separatrix 84. Once they are beyond the separatrix 84, they can leave along the open field lines 80 provided that they experience scattering from ionxe2x80x94ion collisions. Although this collisional process does not lead to diffusion, it can change the direction of the ion velocity vector such that it points parallel to the magnetic field. These open field lines 80 connect the FRC topology of the core with the uniform applied field provided outside the FRC topology. Product ions emerge on different field lines, which they follow with a distribution of energies; advantageously in the form of a rotating annular beam. In the strong magnetic fields found outside the separatrix 84 (typically around 100 kG), the product ions have an associated distribution of gyro-radii that varies from a minimum value of about 1 cm to a maximum of around 3 cm for the most energetic product ions. Initially the product ions have longitudinal as well as rotational energy characterized by xc2xd M(vpar)2 and xc2xd M(vperp)2. vperp is the azimuthal velocity associated with rotation around a field line as the orbital center. Since the field lines spread out after leaving the vicinity of the FRC topology, the rotational energy tends to decrease while the total energy remains constant. This is a consequence of the adiabatic invariance of the magnetic moment of the product ions. It is well known in the art that charged particles orbiting in a magnetic field have a magnetic moment associated with their motion. In the case of particles moving along a slow changing magnetic field, there also exists an adiabatic invariant of the motion described by xc2xd M(vperp)2/B. The product ions orbiting around their respective field lines have a magnetic moment and such an adiabatic invariant associated with their motion. Since B decreases by a factor of about 10 (indicated by the spreading of the field lines), it follows that vperp will likewise decrease by about 3.2. Thus, by the time the product ions arrive at the uniform field region their rotational energy would be less than 5% of their total energy; in other words almost all the energy is in the longitudinal component. While the invention is susceptible to various modifications and alternative forms, a specific example thereof has been shown in the drawings and is herein described in detail. It should be understood, however, that the invention is not to be limited to the particular form disclosed, but to the contrary, the invention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the appended claims. |
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063234998 | claims | 1. An electron beam exposure apparatus comprising: a source for emitting an electron beam; an electron optical system for bringing the electron beam into a focus on a target exposure surface; and electron density distribution adjustment means for adjusting an electron density distribution of the electron beam when the electron beam passes through a pupil plane of said electron optical system or a position conjugate to the pupil plane, in which density distribution an electron density at a peripheral portion on the pupil plane becomes higher than that at a central portion. a source for emitting an electron beam; an electron optical system for bringing the electron beam into a focus on a target surface; and a hollow beam forming stop arranged on the pupil plane of said electron optical system or at a position conjugate to the pupil plane, wherein the electron density distribution at a peripheral portion on the pupil plane becomes higher than that at the central portion. a source for emitting an electron beam; an electron optical system for bringing the electron beam into a focus on a target exposure surface; and an aperture stop for shielding the electron beam entering a center thereof, wherein the electron density distribution at a peripheral portion on the pupil plane becomes higher than that at the central portion. 2. The apparatus according to claim 1, wherein said electron optical system projects, onto said target exposure surface, an electron beam transmitted through a mask held between said source and said electron optical system. 3. The apparatus according to claim 1, wherein said electron optical system projects, onto said target exposure surface, a variable rectangular beam formed by a first aperture having a regular shape and a second aperture having a rectangular shape. 4. The apparatus according to claim 1, wherein said electron density distribution adjustment means includes a stop for forming a hollow beam. 5. The apparatus according to claim 1, further comprising a correction electron optical system arranged between said source and said electron optical system to form a plurality of intermediate images of said source for correcting an aberration generated by said electron optical system, the intermediate images being projected on said target exposure surface by said electron optical system. 6. The apparatus according to claim 5, wherein said correction electron optical system has a plurality of element electron optical systems each of which forms one intermediate image. 7. The apparatus according to claim 6, wherein said electron density distribution adjustment means has a plurality of aperture stops each of which shields an electron beam which enters near an optical axis of a corresponding one of said element electron optical systems. 8. The apparatus according to claim 6, wherein said electron density distribution means is arranged on the pupil plane of said electron optical system or at a position conjugate to the pupil plane. 9. A device manufacturing method of manufacturing a device by using an electron beam exposure apparatus of claim 1. 10. An electron beam exposure apparatus comprising: 11. The apparatus according to claim 10, wherein an electron density distribution of the electron beam is adjusted by said hollow beam forming stop such that an electron density at a peripheral portion on the pupil plane becomes higher than that at a central portion. 12. The apparatus according to claim 10, wherein said electron optical system projects, onto said target exposure surface, an electron beam transmitted through a mask held between said source and said electron optical system. 13. The apparatus according to claim 10, wherein said electron optical system projects, onto said target exposure surface, a variable rectangular beam formed by a first aperture having a rectangular shape and a second aperture having a rectangular shape. 14. A device manufacturing method of manufacturing a device by using an electron beam exposure apparatus of claim 10. 15. An electron beam exposure apparatus comprising: 16. The apparatus according to claim 10, wherein an electron density distribution of the electron beam is adjusted when the electron beam passes through a pupil plane of said electron optical system or a position conjugate to the pupil plane, in which density distribution an electron density at a peripheral portion on a pupil plane of said electron optical system becomes higher than that at a central portion. 17. The apparatus according to claim 15, wherein said aperture stop comprises a stop for forming a hollow beam. 18. The apparatus according to claim 15, wherein said electron optical system projects, onto said target exposure surface, an electron beam transmitted through a mask held between said source and said electron optical system. 19. The apparatus according to claim 15, wherein said electron optical system projects, onto said target exposure surface, a variable rectangular beam formed by a first aperture having a rectangular shape and a second aperture having a rectangular shape. 20. The apparatus according to claim 15, further comprising a correction electron optical system arranged between said source and said electron optical system to form a plurality of intermediate images of said source for correcting an aberration generated said electron optical system, the intermediate images being projected on said target exposure surface by said electron optical system. 21. The apparatus according to claim 20, wherein said correction electron optical system has a plurality of element electron optical systems each of which forms one intermediate image. 22. The apparatus according to claim 20, wherein said electron density distribution adjustment means has a plurality of aperture stops each of which shields an electron beam which enters near an optical axis of a corresponding one of said element electron optical systems. 23. A device manufacturing method of manufacturing a device by using an electron beam exposure apparatus of claim 15. |
description | This application claims the benefit under 35 U.S.C. § 119(e) of U.S. Provisional Patent Application Ser. No. 62/545,340, filed Aug. 14, 2017, the disclosure of which is hereby incorporated herein in its entirety by this reference. The disclosure, in various embodiments, relates generally to assemblies for shielding (e.g., protecting) an underlying structure, such as, for example, an aircraft or a spacecraft (e.g., aerospace structure), to systems including protective assemblies, and to methods of forming protective assemblies and systems. Aerospace vehicles, such as aircraft and spacecraft, may have an external protection system to endure launch, in-flight, and space environments. These environments may subject the aerospace vehicle to impact with foreign objects, such as rain, birds, rocks, dirt, micro-meteoroids, and other orbital debris, that may damage the vehicle. In addition, these foreign objects may approach the aerospace vehicle at subsonic velocities, such about 0.3 km/second, and at hypersonic velocities, such as at least 3 km/second. Furthermore, the aerospace vehicle may also be subject to aerodynamic heat from atmospheric friction generated during launch and flight. The aerospace vehicle may also be subject to heating created by thermal flash, solar and nuclear burst induced radiation, so-called Advanced Threat radiation as may be produced by directed-energy weapons, and the like encountered by the vehicle during flight. Additionally, the aerospace vehicle may also be subject to damage by electromagnetic phenomena, such as electromagnetic pulses and lightning strikes. Adhesive bonding has been previously used to join protection systems to the aerospace vehicle. Such adhesives are provided at an interface between the protection system and the structural body of the aerospace vehicle. Using adhesives that are discrete from either the protection system or the structural body of the aerospace vehicle provides a distinct bond line between the protective system and vehicle skin. Furthermore, adhesives provide a mechanical bond between the protection system and the exterior of the aerospace vehicle rather than a direct chemical bond between the protection system and the exterior of the aerospace vehicle. Consequently, adhesives are a relatively weak bonding method. Accordingly, during operation of the vehicle, the protection system may at least partially break away from the aerospace vehicle and leave a portion of the vehicle exterior exposed to the harmful operating environments previously described. Such adhesives also add weight to the aerospace vehicle, introduce additional mechanical impedance, thermal conductance, and shock propagation, have limited thermal properties, and are difficult to repair in the event of damage. Using adhesive bonding to join layers of the protective system has similar disadvantages to adhesives provided at the interface between the protection system and the structural body of the aerospace vehicle. For example, such adhesive bonding adds weight to the aerospace vehicle, forms a weak mechanical rather than chemical bond, and renders the distinct bond lines susceptible to failure by thermal and mechanical stresses. In some embodiments, the disclosure includes a protective assembly comprising a first region formulated and configured to provide protection from alpha, beta, and electromagnetic radiation and comprising a composite of particles and polymer; a second region formulated and configured to provide protection from ballistic impact and comprising a composite of fibers and polymer; and a third region formulated and configured to provide protection from thermal radiation and comprising a composite of particles, fibers, and polymer. Each of the first region, the second region, and the third region comprises a common polymer. In some embodiments, the disclosure includes an aerospace structure comprising an aerospace structure body and a protective assembly formed on an exterior surface of the aerospace structure body. The protective assembly comprises a first region formulated and configured to provide protection from alpha, beta, and electromagnetic radiation and comprising a composite of particles and polymer; a second region formulated and configured to provide protection from ballistic impact and comprising a composite of fibers and polymer; and a third region formulated and configured to provide protection from thermal radiation and comprising a composite of particles, fibers, and polymer. Each of the first region, the second region, and the third region comprises a common polymer. In some embodiments, the disclosure includes a method of forming a protective assembly for an aerospace vehicle comprising mounting a protective assembly over an aerospace structure body. The protective assembly comprises a first region formulated and configured to provide protection from alpha, beta, and electromagnetic radiation and comprising a composite of particles and polymer; a second region formulated and configured to provide protection from ballistic impact and comprising a composite of fibers and polymer; and a third region formulated and configured to provide protection from thermal radiation and comprising a composite of particles, fibers, and polymer. Each of the first region, the second region, and the third region is provided in an at least partially uncured state. The method further comprises co-curing the protective assembly to form a chemical bond between adjacent regions of the first region, the second region, and the third region and to attach the protective assembly to the aerospace structure body. The illustrations presented herein are not actual views of any particular structure, device, assembly, protective structure (e.g., shield), or aerospace vehicle, but are merely idealized representations employed to describe example embodiments of the disclosure. The following description provides specific details of embodiments of the disclosure in order to provide a thorough description thereof. However, a person of ordinary skill in the art will understand that the embodiments of the disclosure may be practiced without employing many such specific details. Indeed, the embodiments of the disclosure may be practiced in conjunction with conventional techniques employed in the industry. In addition, the description provided below does not include all elements to form a complete structure or assembly. Only those process acts and structures necessary to understand the embodiments of the disclosure are described in detail below. Additional conventional acts and structures may be used. Also note, any drawings accompanying the application are for illustrative purposes only, and are thus not drawn to scale. Additionally, elements common between figures may have corresponding numerical designations. As used herein, the terms “comprising,” “including,” “containing,” “characterized by,” and grammatical equivalents thereof are inclusive or open-ended terms that do not exclude additional, unrecited elements or method steps, but also include the more restrictive terms “consisting of” and “consisting essentially of” and grammatical equivalents thereof. As used herein, the term “may” with respect to a material, structure, feature, or method act indicates that such is contemplated for use in implementation of an embodiment of the disclosure, and such term is used in preference to the more restrictive term “is” so as to avoid any implication that other compatible materials, structures, features and methods usable in combination therewith should or must be excluded. As used herein, the term “configured” refers to a size, shape, material composition, and arrangement of one or more of at least one structure and at least one apparatus facilitating operation of one or more of the structure and the apparatus in a predetermined way. As used herein, the singular forms following “a,” “an,” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. As used herein, spatially relative terms, such as “beneath,” “below,” “lower,” “bottom,” “above,” “upper,” “top,” “front,” “rear,” “left,” “right,” and the like, may be used for ease of description to describe one element's or feature's relationship to another element(s) or feature(s) as illustrated in the figures. Unless otherwise specified, the spatially relative terms are intended to encompass different orientations of the materials in addition to the orientation depicted in the figures. As used herein, the term “substantially” in reference to a given parameter, property, or condition means and includes to a degree that one of ordinary skill in the art would understand that the given parameter, property, or condition is met with a degree of variance, such as within acceptable manufacturing tolerances. By way of example, depending on the particular parameter, property, or condition that is substantially met, the parameter, property, or condition may be at least 90.0% met, at least 95.0% met, at least 99.0% met, or even at least 99.9% met. As used herein, the term “about” used in reference to a given parameter is inclusive of the stated value and has the meaning dictated by the context (e.g., it includes the degree of error associated with measurement of the given parameter). Composites or composite materials as discussed herein may include one or more plies of material or ply layers included a reinforcing material (e.g., reinforcing fibers) in at least some matrix material, which plies may be laid up on a tool one at a time and/or multiple layers at the same time. The plies or layers can be made of any materials with fibers (or plies) that exhibit desired characteristics including, but not limited to, pre-preg material and/or dry fiber material. The pre-preg material and the dry fiber material can include, but are not limited to, unidirectional tapes, bias tapes, woven fabrics, non-woven fabrics, and non-crimp fabrics. The orientation of the fibers (or plies) may also vary throughout the structure. According to embodiments of the disclosure, a protective assembly 100 may include a plurality of regions (e.g., two or more regions) that address multi-functional aspects for protecting an aerospace structure 110. In some embodiments, the aerospace structure 110 may comprise a body of a vehicle configured for flight such as an aircraft or a spacecraft. As used herein, the term “spacecraft” may mean and include vehicles or devices designed for travel or operation outside the Earth's atmosphere. As used herein, the term “aircraft” may mean and include vehicles or devices designed for travel or operation inside the earth's atmosphere. Structure 110 according to embodiments of the disclosure may include a satellite, a missile, including a Ground Based Strategic Deterrent (GBSD), a Ground Based Midcourse Defense (GMD) system, an Exoatmospheric Kill Vehicle (EKV), and other movable or stationary structures. As used herein, the term “aerospace structure” is used to collectively refer to aircraft, spacecraft, satellite, missiles, or other movable or stationary structures for use in Earth's atmosphere and/or surrounding space outside Earth's atmosphere. The protective assembly 100 may be provided on the aerospace structure 110 to protect the aerospace structure 110 against one or more threats that may be encountered by the aerospace structure 110 (e.g., during use) inside and/or outside the Earth's atmosphere. For example, the aerospace structure 110 may be subject to harmful radiation, thermal flash, lightning strikes (e.g., electrostatic pulses), ballistic impact, aerothermal heating, electromagnetic pulses, and/or other harmful environment impacts. The composition of the protective assembly 100 may be selected to provide protection against one or more anticipated threats. According to embodiments of the disclosure and as illustrated in FIGS. 1, 2, and 6, the protective assembly 100 may include a radiation protection region 102, a ballistic protection region 104, an electrical charge protection region 106, and a thermal radiation protection region 108. The electrical charge protection region 106 may protect the aerospace structure 110 from at least one of electrostatic pulses and electromagnetic pulses (EMPs). In some embodiments, the protective assembly may optionally comprise two electrical charge protection regions 106. In such embodiment, a first electrical charge protection region 106a may provide protection against electrostatic pulses, and a second electrical charge protection region 106b may provide protection against EMPs. In further embodiments, the protective assembly 100 may include only those regions that protect against threats with which the aerospace structure 110 is anticipated to encounter. Accordingly, if it is not anticipated that the aerospace structure 110 will encounter a particular threat, protection against this particular threat may not be provided so as to reduce weight and cost of the aircraft or spacecraft. For example, as illustrated in FIG. 3, the protective assembly 100 may lack the radiation protection region 102. Additionally, the regions of the protective assembly 100 may be provided in an order over the aerospace structure 110 depending on the order in which the aerospace structure 110 is expected to encounter one or more threats. Accordingly, the regions of the protective assembly 100 may be provided in an order different from that illustrated in FIGS. 1 and 2. For example, as illustrated in FIG. 1, the protective assembly 100 may comprise from inboard to outboard, the radiation protection region 102, the second electrical charge protection region 106b, the ballistic protection region 104, the first electrical charge protection region 106a, and the thermal radiation protection region 108. In other embodiments, as illustrated in FIG. 3, the protective assembly 100 includes from inboard to outboard the ballistic protection region 104, the electrical charge protection region 106, and the thermal radiation protection region 108 from inboard to outboard. In FIG. 4, the protective assembly 100 includes from inboard to outboard the ballistic protection region 104, the thermal radiation protection region 108, and the electrical charge protection region 106. In some embodiments, one or more of the various regions may be provided with multiple layers or sheets of material. In some embodiments, one or more of the various regions may be provided in a same region or volume of material. The radiation protection region 102 may be configured to protect the aerospace structure 110 from radiation due to radioactivity and/or electromagnetic radiation. For example, the radiation protection region 102 may be configured to protect the underlying aerospace structure 110 from one or more of X-ray radiation, gamma radiation, alpha radiation, and beta radiation. The radiation protection region 102 may comprise a particle-matrix composite material including a matrix 112 having a plurality of particles 114 dispersed therein (FIG. 6). As used herein, the term “particles” may mean and include any coherent volume of solid matter having an average dimension of 1 mm or less. In some embodiments, the particles 114 may comprise microparticles, or particles having an average particle diameter of about 10 μm or less. In other embodiments, the particles 114 may comprise nanoparticles, or particles having an average particle diameter about 100 nm or less. In yet other embodiments, the particles 114 may comprise a mixture of microparticles and nanoparticles. The particles 114 may comprise a metal or metal alloy and, more particularly, a metal having a high atomic number (e.g., a high Z metal) or an alloy comprising such a metal. The metal or metal alloy of the particles 114 may be selected based on the efficiency of the metal or metal alloy to at least partially (e.g., substantially entirely, entirely) prevent radiation from passing through the radiation protection region 102 as the particles 114 reflect radiation to which the aerospace structure 110 and assembly 100 may be exposed. In some embodiments, the particles 114 may comprise a metal or metal alloy having an atomic number greater than or equal to 26. By way of non-limiting example, the particles 114 may comprise at least one of iron (Z=26), tantalum (Z=73), tungsten (Z=74), bismuth (Z=83), or uranium (Z=92). The radiation protection region 102 may comprise up to about 80 weight percent (wt %), up to 85 wt %, up to 90 wt %. The matrix 112 of the radiation protection region 102 may comprise a polymeric material such that the radiation protection region 102 comprises a particle-polymer composite. In some embodiments, the polymeric material may be an elastomeric material. The polymeric material may be selected to comprise an ethylene acrylic elastomer (AEM), such as VAMAC®, rubber, such as a silicone rubber or an ethylene propylene diene monomer (EPDM) rubber, or a styrene block-based thermoplastic elastomer. The ballistic protection region 104 may be configured to protect the aerospace structure 110 from impact with foreign objects during use of the structure. For example, the ballistic protection region 104 may be configured to protect the aerospace structure 110 from hail, rocks, dirt, micro-meteoroids, and other debris that may exist inside and outside of Earth's atmosphere as well as erosion from dust and/or rain. The ballistic protection region 104 may comprise a composite material including fibers 116 embedded with or within a matrix 118. The fibers 116 may comprise aramid fibers, such as para-aramid synthetic fibers (e.g., KEVLAR®), carbon fibers, and/or boron fibers. In some embodiments, the fibers 116 may form a dispersed phase within the matrix 118, which may form a substantially continuous matrix phase. The fibers 116 may be formed of at least one of plies of fibers (e.g., woven fiber fabric) or unidirectional fiber windings. In some embodiments, the fibers 116 may be oriented within the matrix 118 to protect the aerospace structure 110 from impact with foreign objects traveling in an anticipation direction of impact with the assembly 100 and the aerospace structure 110. In such embodiments, a portion of the fibers 116 may be oriented in a direction substantially parallel with the anticipated direction of impact. Another portion of the fibers 116 may be oriented in a direction substantially perpendicular with the anticipated direction of impact in order to prevent displacement of the fibers 116 oriented parallel to the anticipated direction of impact. Such selective orientation of the fibers 116 within the matrix 118 enables the amount of fibers 116 included within the ballistic protection region 104 to effectively protect the aerospace structure 110 can be reduced by, for example, eliminating randomly oriented fibers and, as a result, the weight of the ballistic protection region 104 and the protective assembly 100 as a whole may be reduced. In some embodiments, the ballistic protection region 104 comprises a plurality of layers of fibers embedded with or within the matrix 118, such as a plurality of embedded fabric layers, as illustrated in FIGS. 6 and 7. The number of fiber layers may be selected depending on the impact resistance necessary during use of the aerospace structure 110. The composite of fibers 116 and matrix 118 may have a sufficient impact resistance to resist (e.g., inhibit, prevent) penetration by and absorb energy from any projectile that may impact the aerospace structure 110 and the assembly 100. In some embodiments, the ballistic protection region 104 may further comprise fiber pulp to provide the underlying aerospace structure 110 with additional ballistic and thermal protection. The ballistic protection region 104 may comprise up to about 20 parts per hundred of the fiber pulp in the matrix 118. The matrix 118 may comprise a polymeric material. Accordingly, the ballistic protection region 104 may comprise a fiber-reinforced polymer material. The polymeric material may comprise an elastomeric material. In some embodiments, the matrix 118 may comprise an ethylene acrylic elastomer, such as VAMAC®, rubber, such as a silicone rubber or an ethylene propylene diene monomer (EPDM) rubber, or a styrene block-based thermoplastic elastomer. The electrical charge protection region 106 may be configured to protect the aerospace structure 110 from at least one of electrostatic discharge (e.g., electrostatic pulses) and electromagnetic discharge (e.g., electromagnetic discharge). The electrical charge protection region 106 may comprise a metallic material 120 of a metal or metal alloy. The metal or metal alloy may be highly electrically conductive and/or may have high magnetic permeability. In some embodiments, the metallic material 120 may be porous and may comprise a metallic mesh. In such embodiments, the electrical charge protection region 106 may further comprise a matrix 107 (FIG. 6) that permeates pores (e.g., openings) within the mesh of the metallic material 120. The matrix 107 may comprise a polymeric material such as an elastomeric material. The elastomeric material may be selected to comprise an ethylene acrylic elastomer, such as such as VAMAC®, rubber, such as a silicone rubber or an ethylene propylene diene monomer (EPDM) rubber, or a styrene block-based thermoplastic elastomer. In other embodiments, the metallic material 120 may be non-porous and may comprise a metallic foil. In operation, the mesh or foil of the metallic material 120 may dissipate the pulses of energy through the conductive pathways thereof to and at least partially prevent the energy from passing to the underlying aerospace structure 110. In some embodiments, the mesh or foil 120 may comprise silver, copper, aluminum, nickel, and iron, as well as alloys of and combinations of the foregoing. In some embodiments, separate electrical charge protection regions may be provided that provide protect the aerospace structure 110 from electrostatic pulses and electromagnetic pulses, respectively. As illustrated in FIG. 1, in such embodiments, the protective assembly 100 may comprise the first electrical charge protection region 106a providing protection against electrostatic pulses and the second electrical charge protection region 106b providing protection against EMPs. The first and second electrical charge protection regions 106a, 106b are illustrated in FIG. 1 having a dashed outline to indicate that one or both of the first and second electrical charge protection regions 106a, 106b may be optionally provided. The electrical charge protection region 106a may comprise a mesh of a highly conductive metal or metal alloy. The mesh of the electrical charge protection region 106a may comprise copper, aluminum, and silver, as well as alloys and combinations of the foregoing. The electrical charge protection region 106b may comprise a foil of a highly conductive material and a foil of a material exhibiting high magnetic permeability. In such embodiments, the electrical charge protection region 106b may comprise at least one foil of copper, aluminum, and silver, as well as alloys and combinations thereof and at least one foil of iron and nickel, as well as alloys and combinations thereof. Such foils may be coupled (e.g., laminated) together to form the electrical charge protection region 106b according to ASTM Standard D2651 (Standard Guide for Preparation of Metal Surfaces for Adhesive Bonding). The foils may be coupled (e.g., adhesively bonded) to each other and to an adjacent region of the protective assembly 100 by an adhesive 122. The adhesive 122 may comprise a polymeric material. In some embodiments, the adhesive 122 may be selected to comprise at least one of an epoxy, a bi-maleimide (BMI) resin, and an ethylene acrylic elastomer, such as VAMAC®, rubber, such as a silicone rubber or an ethylene propylene diene monomer (EPDM) rubber, or a styrene block-based thermoplastic elastomer. The thermal radiation protection region 108 may be configured to protect the aerospace structure 110 against thermal radiation. In some embodiments, the thermal radiation protection region 108 may further be configured to protect the aerospace structure 110 against rain ablation. The thermal radiation protection region 108 may comprise a composite material including a matrix 124 and a dispersed phase. In some embodiments, the matrix 124 (FIG. 6) of the composite material may comprise a polymeric material such as an elastomeric material. The elastomeric material may comprise an ethylene acrylic elastomer, such as VAMAC®, rubber, such as a silicone rubber or an ethylene propylene diene monomer (EPDM) rubber, or a styrene block-based thermoplastic elastomer. In some embodiments, the dispersed phase may comprise particles 126 (FIG. 6) dispersed in the matrix 124. The particles 126 may be white in appearance and may provide the thermal radiation protection region 108 generally with a white coloring. In some embodiments, the particles 126 may comprise boron nitride or titanium dioxide. Alternatively or additionally, the thermal radiation protection region 108 may comprise a dispersed phase of fibers other particles, fibers, and the like to lower the thermal diffusivity of the thermal radiation protection region 108. In such embodiments, the dispersed phase of the thermal radiation protection region 108 may comprise one or more of para-aramid synthetic fibers (e.g., KEVLAR®), para-aramid fiber pulp, para-aramid fiber rovings, and hollow, ceramic microspheres. While each region of the assembly 100 has been described with regard to a primary or individual protective function, the assembly 100 as a whole may be configured to protect the aerospace structure 110 against thermal damage. The aerospace structure 110 may be subject to heat produced from thermal radiation, from the high-speed passage of the aerospace structure 110 through air, and the like. In some embodiments, the assembly 100 may be configured as a thermal insulator and may reduce the amount of heat transferred to the aerospace structure 110. For instance, inclusion of a dispersed phase comprising, for example, para-aramid fibers, para-aramid pulp, and/or ceramic microspheres as previously described herein to one or more of the regions of the assembly 100 may reduce the thermal conductivity and/or may increase ablation resistance of one or more regions of the assembly 100. The thermal conductivity and density of one or more regions of the assembly 100 may be reduced by introducing blow agents, foaming agents, accelerators, or other mechanisms to provide gas pockets in the polymeric matrix and form a foam matrix. The foam may include open cells and/or closed cells. In some embodiments, the foam may have up to about 50% porosity. Furthermore, the matrix of two or more (e.g., each) regions of the assembly 100 may comprise a common (e.g., substantially the same) polymeric material. For example, the matrix of two or more of the regions (e.g., adjacent regions) of the assembly 100 may be similar in composition (e.g., 80% similar in composition, 90%, 95%, 100%). In some embodiments, a majority of (e.g., an entirety of) the matrix of two or more of the regions of the assembly 100 may be similar in composition. In some embodiments, at least a portion of the matrix at a common boundary or interface between two or more adjacent regions of the assembly 100 may be similar in composition. For example, one or more of the matrices of the regions may gradually transition to a common material at the interface. By selecting each region or at least adjacent regions of the assembly 100 to comprise a common polymeric material, an interface between adjacent regions of the assembly 100 may be impedance matched and/or thermal expansion matched (e.g., having substantially the same coefficient of thermal expansion). In operation, interfaces that are impedance and thermally matched are less susceptible to separating (e.g., delaminating, failing) due to interactions with the environment in which the assembly 100 is exposed during use as compared to interfaces between dissimilar materials. In some embodiments, the aerospace structure 110 over which the assembly 100 is applied may be formed of a composite material that includes a reinforcing phase (e.g., dispersed phase) disposed within a matrix. For example, the reinforcing phase may comprise particles, whiskers, fibers, etc. In some embodiments, the reinforcing phase may comprise a carbon fiber reinforcing phase or may include other reinforcing fibers, such as, for example, fiberglass, aramids, boron, ceramics, or combinations thereof. The matrix may comprise a polymeric material. The polymeric material may comprise an elastomeric polymer, such as a cyanate ester polymer matrix. In other embodiments, the aerospace structure 110 may comprise a metal or metal alloy. Embodiments of the disclosure also relate to methods of forming the protective assembly 100 and, more particularly, to forming the protective assembly 100 on the aerospace structure 110. FIG. 5 illustrates a cross-sectional view of the protective assembly 100 enclosed about the aerospace structure 110. As illustrated in FIG. 4, the aerospace structure 110 is a cylindrical portion of an aerospace structure body 111. The assembly 100 may be formed by providing (e.g., overlying) each of the respective layers of each of the radiation protection region 102, the ballistic protection region 104, the electrical charge protection region 106, and the thermal radiation protection region 108 over the aerospace structure body 111 of the aerospace structure 110. At least a portion of one or more (e.g., all) of the foregoing regions of the assembly 100 is provided over the aerospac structure body 111 of the aerospace structure 110 in an at least partially uncured state (e.g., entirely uncured, partially uncured, a majority uncured). More particularly, the polymeric material of each of the foregoing regions of the protective assembly may be initially provided in an uncured state. The matrix of the aerospace structure 110 may also be provided in an at least partially uncured state. In some embodiments, the layers of one or more of the foregoing regions may be provided over the aerospace structure 110 by a wet lay-up process. In other embodiments, layers of one or more of the foregoing regions may be provided by a filament winding process or hand layup of pre-preg. The assembly 100 may be bonded to the aerospace structure 110 by a co-curing process. As used herein, the term “co-curing” may mean and include a curing process in which two or more materials each in an at least partially uncured state (e.g., entirely uncured, partially uncured, a majority uncured) are joined and bonded together during the same process. In the co-curing process, the uncured regions of the protective assembly 100 may be joined together by simultaneously curing and bonding. Such curing and bonding may result in a chemical bond and/or a mechanical interlocking at the interface between the respective regions of the assembly 100. As previously described herein, the matrix of one or more regions of the assembly 100 is substantially the same polymeric material and, as a result of co-curing the common matrix, the interface between adjacent regions of the assembly 100 may be substantially indistinct (e.g., seamless) as illustrated in FIG. 7. Further, the matrix of the one or more regions may be continuous and chemically bonded (e.g., intermingled) at the interface. Unlike a co-bonding process, or a curing process in which two or more materials at least one of which is in a fully cured are joined and bonded together, or an adhesive bonding process, or a process in which two or more materials are joined together by a curing process of an adhesive therebetween, the co-curing process results in a chemical bond between the regions of the protective assembly 100 and between the protective assembly 100 and the aerospace structure 110. Accordingly, the regions of the protective assembly 100 have higher interfacial strength and are less susceptible to separating (e.g., delaminating, failing) due to interactions with the environment in which the assembly 100 is exposed during use compared to structures formed by co-bonding or adhesive bonding. In some embodiment, the matrix of the regions of the protective assembly 100 may be co-cured with the aerospace structure 110, which may be provided in an at least partially uncured state. In other embodiments, the matrix of the regions of the protective assembly 100 may be co-cured and the protective assembly 100 may be co-bonded to the aerospace structure. In yet other embodiments, such as embodiments in which the aerospace structure 110 comprises a metal or metal alloy, the assembly 100 may be bonded to the aerospace structure 110 by an adhesive bonding or vulcanization process. During the co-curing process, heat is applied to cure the polymeric material of one or more regions of the assembly 100 and to bond the regions together at their respective interfaces. In some embodiments, the aerospace structure 110 and the assembly 100 may be heated in an auto-clave or an oven for the co-curing process. In some embodiments, the auto-clave or oven may heat the aerospace structure 110 and the assembly 100 to a temperature of up to about 400° F. In some embodiments, force (e.g., pressure) may be applied to the aerospace structure 110 and the assembly 100 during the co-curing process. As illustrated in FIG. 5, in such embodiments, the pressure may be provided by a film 128 provided about the exterior surface of the assembly 100. The film 128 may apply a compressive pressure to the aerospace structure 110 and the assembly 100 throughout the co-curing process while allowing gases produced during the co-curing that may cause pitting or other deformations of the aerospace structure 110 and/or assembly 100 to escape through the film 128. In some embodiments, the film 128 may comprise a fiberglass wrap, such as Armalon (a polytetrafluoroethylene coated fiberglass). In other embodiments, the film 128 may comprise a vacuum bag. In such embodiments, a vacuum may be drawn on the bag to remove air and other gases produced during the co-curing process from within the bag. While the disclosed structures and methods are susceptible to various modifications and alternative forms in implementation thereof, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the disclosure is not limited to the particular forms disclosed. Rather, the present invention encompasses all modifications, combinations, equivalents, variations, and alternatives falling within the scope of the disclosure as defined by the following appended claims and their legal equivalents. |
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description | Priority is claimed to Japanese Patent Application No. 2016-171800, filed Sep. 2, 2016, the entire content of which is incorporated herein by reference. A certain embodiment of the present invention relates to a charged particle beam therapy apparatus and a ridge filter. Charged particle beam therapy apparatuses performing irradiation of a charged particle beam are known. There are cases where such a charged particle beam therapy apparatus has a ridge filter for generating a spread out Bragg peak of the charged particle beam (for example, refer to the related art). The ridge filter has a plurality of damping portions which are extended and individually have a triangular cross-sectional shape. A support section lying throughout the whole region of the ridge filter supports a bottom portion side of the plurality of damping portions. Energy of charged particles configuring a charged particle beam is degraded in accordance with the thickness of the damping portion to passthrough. Therefore, in an irradiating direction of the charged particle beam, the energy of charged particles passing through a thick part of the damping portion is low, and the energy of charged particles passing through a thin part of the damping portion remains high. As a result, it is possible to obtain a charged particle beam in which charged particles having different energy are mixed together, that is, a charged particle beam having energy within a certain degree of width (having a spread out Bragg peak). According to an aspect of the present invention, there is provided a charged particle beam therapy apparatus including an accelerator configured to accelerate a charged particle and to emit a charged particle beam, an irradiation unit configured to irradiate an irradiation subject with the charged particle beam, and a ridge filter that is provided in the irradiation unit and generates a spread out Bragg peak of the charged particle beam. The ridge filter includes a plurality of damping members reducing energy of the incident charged particle beam, in an intersecting direction which intersects an irradiating direction of the charged particle beam. The damping member has a cross-sectional area changing along the irradiating direction and has a side surface in a case of being seen in the intersecting direction, being bonded to a side surface of another damping member. A pass-through portion passing through the ridge filter in the irradiating direction is formed at a position different from a position of the damping member in a case of being seen in the irradiating direction. According to another aspect of the present invention, there is provided a ridge filter for a charged particle beam therapy apparatus generating a spread out Bragg peak of a charged particle beam. The ridge filter includes a plurality of damping members reducing energy of an incident charged particle beam in an intersecting direction which intersects an irradiating direction of the charged particle beam. The damping member has a cross-sectional area changing along the irradiating direction and has a side surface in a case of being seen in the intersecting direction, being bonded to a side surface of another damping member. A pass-through portion passing through the ridge filter in the irradiating direction is formed at a position different from a position of the damping member in a case of being seen in the irradiating direction. In a charged particle beam therapy apparatus of the related art, when a charged particle beam is incident on a ridge filter, the charged particle beam is incident not only on a damping portion but also on a support member. Therefore, there may be a disadvantage in that the charged particle beam scatters when the charged particle beam passes through the support member. If the charged particle beam scatters unexpectedly, there is a possibility that desired dose distribution of the charged particle beam will not be able to be obtained. It is desirable to provide a charged particle beam therapy apparatus in which a charged particle beam can be restrained from scattering, and a ridge filter. In a charged particle beam therapy apparatus according to an embodiment of the present invention, the ridge filter may include a plurality of damping members reducing energy of an incident charged particle beam, in an intersecting direction which intersects an irradiating direction of the charged particle beam. Here, the damping member may have aside surface in a case of being seen in the intersecting direction, being bonded to a side surface of another damping member. In this manner, when the adjacent damping members support each other, even if there is provided no support member supporting all the damping members, it is possible to ensure the strength for serving as the ridge filter. Since the strength can be ensured even if there is provided no support member, it is possible to form a pass-through portion which passes through the ridge filter in the irradiating direction at a position different from a position of the damping member in a case of being seen in the irradiating direction. According to such a structure, a charged particle beam which is not incident on the damping member can pass through the pass-through portion, and thus, the charged particle beam can travel to the downstream side of the ridge filter without scattering. Consequently, the charged particle beam can be restrained from scattering. In the charged particle beam therapy apparatus according to the embodiment of the present invention, the intersecting direction may have a first direction and a second direction which intersects the first direction. The damping members may individually have a pyramid shape, may be arranged along the first direction, and may be arranged along the second direction. According to such a configuration, even if there is provided no support member, it is possible to ensure the strength for serving as the ridge filter. In addition, since the damping members individually have a pyramid shape and are arranged along the first direction and the second direction, the damping members are disposed in a two-dimensional array. For example, in a ridge filter according to a comparative example as illustrated in FIGS. 7A and 7B, the damping members extending straight in the second direction are arranged in the first direction. Accordingly, there appears shade of a Bragg peak in the first direction, and there appears no shade in the second direction, resulting in shade having a striped pattern in a case of being seen in a planar manner. Meanwhile, since the damping members are disposed in a two-dimensional array, it is possible to obtain planar shade of the Bragg peak, so that the approximately even shade of the Bragg peak can be realized. According to the ridge filter of the embodiment of the present invention, it is possible to obtain an operation and an effect similar to those of the charged particle beam therapy apparatus. According to the embodiment of the present invention, the charged particle beam can be restrained from scattering. Hereinafter, with reference to the accompanying drawings, a charged particle beam therapy apparatus and a ridge filter according to an embodiment of the present invention will be described. In the descriptions of the drawings, the same reference sign will be applied to the same element, and the description thereof will not be repeated. FIG. 1 is a sectional view of a structure 50 in which the charged particle beam therapy apparatus according to the embodiment of the present invention is installed. FIG. 2 is a perspective view illustrating a configuration in the vicinity of an irradiation unit of the charged particle beam therapy apparatus according to the embodiment of the present invention. FIG. 3 is a schematic diagram of a configuration of the charged particle beam therapy apparatus according to the embodiment of the present invention. As illustrated in FIG. 3, a charged particle beam therapy apparatus 1 irradiates a tumor (irradiation subject) 14 inside the body of a patient 15 with a charged particle beam R. The charged particle beam R is realized by accelerating electrically charged particles at a high speed. Examples of the charged particle beam R include a proton beam, a heavy particle (heavy ion) beam, and a particle beam. Hereinafter, descriptions will be given by using the terms of “X-axis direction”, “Y-axis direction”, and “Z-axis direction”. The “Z-axis direction” is a direction in which a base axis AX of the charged particle beam R extends inside an irradiation unit 8 (will be described below in detail). The “base axis AX” is an irradiation axis of the charged particle beam R in a case where the charged particle beam R is not deflected due to scanning electromagnets 3a and 3b (will be described below). FIG. 3 illustrates a state where irradiation is performed with the charged particle beam R along the base axis AX. In the descriptions below, a direction in which irradiation is performed with the charged particle beam R along the base axis AX will be considered as “irradiating direction of the charged particle beam R”. The “X-axis direction” is a direction within a plane orthogonal to the Z-axis direction. The “Y-axis direction” is a direction orthogonal to the X-axis direction within a plane orthogonal to the Z-axis direction. As illustrated in FIGS. 1 and 3, the charged particle beam therapy apparatus 1 includes an accelerator 2, a rotary gantry 3 including the irradiation unit (irradiation nozzle) 8, a transportation line 40, and a control device 7. As illustrated in FIG. 1, the charged particle beam therapy apparatus 1 is installed in the structure 50. In the example illustrated in FIG. 1, the structure 50 is a multi-story building (here, a two-story building). The accelerator 2 and the rotary gantry 3 are respectively installed in rooms on each floor. The transportation line 40 is provided across the floors of the structure 50 and connects the accelerator 2 and the irradiation unit 8 of the rotary gantry 3 to each other (refer to FIG. 2). The rotary gantry 3 can rotate or oscillate around a treatment bed 4 on which the patient 15 is placed. The irradiating direction of a charged particle beam emitted by the irradiation unit 8 can be changed by rotating the rotary gantry 3. The accelerator 2 and the rotary gantry 3 may be installed on the same floor instead of being installed on floors different from each other. In addition, the irradiation unit 8 may be in a fixed state inside a room (so-called fixed irradiation method) instead of being attached to the rotary gantry 3. The accelerator 2 accelerates charged particles and emits the charged particle beam R. Examples of the accelerator 2 include a cyclotron, a synchrotron, a cyclo-synchrotron, and a linear accelerator. The charged particle beam R generated by the accelerator 2 is transported to the irradiation unit 8 through the transportation line 40. The accelerator 2 is connected to the control device 7, and the control device 7 controls the operation of the accelerator 2. The transportation line 40 is a line for transporting the charged particle beam R, which is emitted from the accelerator 2, to the irradiation unit 8. The transportation line 40 is provided with a duct of which the inside is in a vacuum state or filled with inert gas, a deflected electromagnet which generates a magnetic field for changing the traveling direction of the charged particle beam R passing through the inside of the duct, and the like. The irradiation unit 8 irradiates the patient 15 placed on the treatment bed 4 inside the rotary gantry 3, with the charged particle beam R. The irradiation unit 8 includes the scanning electromagnets 3a and 3b, monitors 4a and 4b, a scatterer 5, a ridge filter 22, a degrader 30, a multi leaf collimator 24, a bolus 26, and a patient collimator 27. Each of components inside the irradiation unit 8 may be suitably omitted or may retreat to a position where irradiation of the charged particle beam R is not hindered, in accordance with the irradiation method of the charged particle beam R. In addition, the degrader 30 and the patient collimator 27 may be omitted. A pair of electromagnets configures the scanning electromagnets 3a and 3b. A magnetic field between the pair of electromagnets is changed in response to a current supplied from the control device 7, and scanning is performed with the charged particle beam R passing through between the electromagnets. The X-axis directional scanning electromagnet 3a performs scanning with the charged particle beam R in the X-axis direction, and the Y-axis directional scanning electromagnet 3b performs scanning with the charged particle beam R in the Y-axis direction. The scanning electromagnets 3a and 3b are disposed on the base axis AX, that is, in order toward the downstream side of the accelerator 2. The scanning electromagnets 3a and 3b perform scanning with the charged particle beam R such that the charged particle beam R is on a trajectory set in advance (for example, a circular trajectory and a zig-zag trajectory). The monitor 4a monitors the beam position of the charged particle beam R, and the monitor 4b monitors the absolute value of the dose of the charged particle beam R and dose distribution of the charged particle beam R. Each of the monitors 4a and 4b outputs monitored information obtained through the monitoring, to the control device 7. The monitor 4a is disposed on the base axis AX of the charged particle beam R, that is, on the downstream side of the accelerator 2 and on the upstream side of the X-axis directional scanning electromagnet 3a. The monitor 4b is disposed on the base axis AX, that is, on the downstream side of the degrader 30. However, the position of each of the monitors 4a and 4b is not particularly limited. The scatterer 5 causes the charged particle beam R passing through the scatterer 5 to be diffused as a wide beam having a width in a direction orthogonal to the irradiation axis. The scatterer 5 has a plate shape and is formed of tungsten having a thickness of several millimeters, for example. On the base axis AX, the scatterer 5 is disposed on the downstream side of the scanning electromagnet 3b and on the upstream side of the monitor 4b. The ridge filter 22 adjusts the dose distribution of the charged particle beam R. Specifically, the ridge filter 22 applies a spread out Bragg peak (SOBP) to the charged particle beam R such that the spread out Bragg peak copes with the thickness of the tumor 14 (length in the irradiating direction) inside the body of the patient 15. Accordingly, a Bragg peak of the charged particle beam R spreads evenly in the thickness direction (here, the Z-axis direction). The ridge filter 22 is disposed on the downstream side of the scatterer 5 and on the upstream side of the monitor 4b, on the base axis AX. The ridge filter 22 will be described below in detail. The degrader 30 is disposed between the ridge filter 22 and the monitor 4b on the base axis AX. The degrader 30 degrades the energy of the charged particle beam R passing through the degrader 30 and adjusts the range of the charged particle beam R. In regards to the adjustment of the range, a rough adjustment may be performed by a degrader (not illustrated) provided immediately behind the accelerator 2, and a fine adjustment may be performed by the degrader 30 inside the irradiation unit 8. The degrader 30 is provided on the base axis AX, that is, on the downstream side in the charged particle beam R beyond the scanning electromagnets 3a and 3b, the scatterer 5, and the ridge filter 22. The degrader 30 adjusts the maximum arrival depth of the charged particle beam R inside the body of the patient 15. However, the position of the degrader 30 is not particularly limited. The degrader 30 is a plate-shaped member which spreads in the X-axis direction and the Y-axis direction. The multi leaf collimator (hereinafter, will be referred to as “MLC”) 24 forms a shape of the charged particle beam R (planar shape) in a plane direction perpendicular to the irradiating direction and has beam blocking portions 24a and 24b each including a plurality of teeth. The beam blocking portions 24a and 24b are disposed so as to face each other. An opening portion 24c is formed between the beam blocking portions 24a and 24b. The MLC 24 allows the charged particle beam R to pass through the opening portion 24c, thereby cutting out the charged particle beam R along a contour corresponding to the shape of the opening portion 24c. In addition, the MLC 24 causes the beam blocking portions 24a and 24b to move forward and backward in a direction orthogonal to the Z-axis direction, thereby being capable of changing the position and the shape of the opening portion 24c. Moreover, the MLC 24 is guided by a linear guide 28 along the irradiating direction and can move along the Z-axis direction. The MLC 24 is disposed on the downstream side of the monitor 4b. The bolus 26 forms a three-dimensional shape of a part of the charged particle beam R at the maximum arrival depth along the shape of a part of the tumor 14 at the maximum depth. For example, the shape of the bolus 26 is calculated based on the contour line of the tumor 14 and the electron density of surrounding tissue obtained from data of an X-ray CT. The bolus 26 is disposed on the downstream side of the MLC 24, on the base axis AX. The patient collimator 27 finally forms the planar shape of the charged particle beam R along the planar shape of the tumor 14. The patient collimator 27 may be disposed on the downstream side of the bolus 26, on the base axis AX, thereby being used as a substitute for the MLC 24, or both the MLC 24 and the patient collimator 27 may be used. The bolus 26 and the patient collimator 27 are provided at a tip portion 8a of the irradiation unit 8. For example, the control device 7 is configured to include a CPU, a ROM, and a RAM. Based on the monitored information output from the monitors 4a and 4b, the control device 7 controls the operation of each of the accelerator 2, the scanning electromagnets 3a and 3b, the scatterer 5, the ridge filter 22, the degrader 30, and the MLC 24 as necessary. In a case of performing irradiation of the charged particle beam R through a wobbler method (broad beam method) by using the charged particle beam therapy apparatus 1 illustrated in FIG. 3, the beam blocking portions 24a and 24b of the MLC 24 move forward and backward along the shape of the tumor 14, so that the opening portion 24c has a predetermined shape. In addition, the bolus 26 for the patient 15 to be treated and the patient collimator 27 are attached to the irradiation unit 8. Subsequently, the charged particle beam R is emitted from the accelerator 2. The emitted charged particle beam R is used for performing scanning by the scanning electromagnets 3a and 3b so as to draw a predetermined trajectory and is diffused by the scatterer 5. Thereafter, the emitted charged particle beam R is formed and adjusted by the ridge filter 22, the degrader 30, the MLC 24, the bolus 26, and the patient collimator 27. Accordingly, the tumor 14 is irradiated with the charged particle beam R within an even irradiation range along the shape of the tumor 14. FIG. 3 illustrates an example of the charged particle beam therapy apparatus 1 performed through the wobbler method. However, the charged particle beam therapy apparatus 1 may be configured to switch between the wobbler method and a different irradiation method (for example, layer stacking method) by switching the state of setting and retreatment of required configuration elements. In addition, since the ridge filter 22 according to the present embodiment can be used regardless of the irradiation method in a case of generating a spread out Bragg peak, the ridge filter 22 may be used when irradiation is performed through the layer stacking method. In a case of performing irradiation through the layer stacking method, a thin ridge filter 22 is used such that a spread out Bragg peak to be generated becomes smaller than a spread out Bragg peak at the time of the wobbler method (broad beam method). Next, the configuration of the ridge filter 22 will be described in detail by using FIGS. 4 to 6. FIG. 4 is a view of a ridge filter seen from an upstream side toward a downstream side in an irradiating direction of a charged particle beam. FIG. 5A is an enlarged view of the ridge filter. FIG. 5B is an enlarged view of damping members configuring the ridge filter. FIG. 6A is a cross-sectional view taken along line VIA-VIA illustrated in FIG. 5A. FIG. 6B is a cross-sectional view taken along line VIB-VIB illustrated in FIG. 5A. In the description of the present embodiment, the XY-plane direction corresponds to “intersecting direction” in Claim, the X-axis direction corresponds to “first direction”, and the Y-axis direction corresponds to “second direction”. However, any positions in the XY-plane direction may respectively correspond to “the first direction” and “the second direction”. As illustrated in FIG. 4, the ridge filter 22 has a rectangular plate shape in its entirety and includes a main body portion 11 which is configured to have a rectangular plate shape, and frame bodies 12 which form a rectangular ring surrounding the main body portion 11 throughout the entire circumference. The frame bodies 12 are extended respectively along four sides of the main body portion 11 and are interlinked together at four corner portions of the main body portion 11. As described below, the main body portion 11 is a member having a plurality of pass-through portions 16. The frame body 12 can ensure the strength of the ridge filter 22 by supporting four sides of the main body portion 11. As the material of the ridge filter 22, for example, aluminum and resin may be used. The material of the main body portion 11 and the material of the frame body 12 may be the same as each other, or the materials may be different from each other. In a case of employing resin as the material of the ridge filter 22, it is possible to form a three-dimensional structure of the ridge filter 22 by using an optical shaping apparatus. As illustrated in FIG. 5A, the main body portion 11 of the ridge filter 22 includes a plurality of damping members 13 reducing the energy of the incident charged particle beam R. The damping members 13 are arranged on the XY-plane which intersects the irradiating direction of the charged particle beam R (direction in which the base axis AX extends). The damping member 13 is a member of which a cross-sectional area changes along the irradiating direction of the charged particle beam R. In the present embodiment, as illustrated in FIG. 5B, the damping member 13 has a pyramid shape, that is, a quadrangular pyramid, of which the bottom surface is a substantial square (octagon of which four corners in the positive direction are chamfered), as a basic shape. Therefore, the cross-sectional area of the damping member 13 increases in the irradiating direction from the upstream side toward the downstream side (Z-axis positive direction). FIG. 5B is a perspective view illustrating the shape of each of the plurality of damping members 13 in the main body portion 11. As described below, as a surface bonded to another damping member 13, the damping member 13 has a side surface 13c on which a bottom surface 13a and two inclined surfaces 13b adjacent to each other are cut such that the corner portion is formed into a plane. The side surface 13c has a triangle shape. Here, the side surface 13c is defined for the convenience of description. However, in the actual ridge filter 22, since the damping members 13 are integrally bonded to each other, the boundary surface between the side surfaces 13c is in a vanished state. As illustrated in FIG. 5A, in the present embodiment, one diagonal line on the bottom surface 13a of the damping member 13 is parallel to the Y-axis direction, the other diagonal line is disposed so as to be parallel to the X-axis direction. In such a state, the plurality of damping members 13 are arranged in the Y-axis direction, and the plurality thereof are arranged in the X-axis direction. That is, the damping members 13 are disposed in a two-dimensional manner along at least two directions. The damping member 13 has the side surfaces 13c on both end sides in the X-axis direction and has the side surfaces 13c on both end sides in the Y-axis direction. Therefore, the damping members 13 adjacent to each other in the X-axis direction are bonded to each other via the side surfaces 13c. The damping members 13 adjacent to each other in the Y-axis direction are bonded to each other via the side surfaces 13c. In addition, when the downstream side is seen from the upstream side in the irradiating direction, oblique sides 13d of each damping member 13 are interlinked with each other and are arranged so as to form a straight line extending straight in the Y-axis direction and a straight line extending straight in the X-axis direction. According to such a configuration, the side surface 13c of the damping member 13 in a case of being seen in the XY-plane direction is bonded to the side surface 13c of another damping member 13. In the present embodiment, as illustrated in FIG. 6A, the side surface 13c of the damping member 13 in a case of being seen in the Y-axis direction is bonded to the side surface 13c of another damping member 13. In addition, on the upstream side in the irradiating direction (Z-axis directional negative side), the oblique sides 13d of each damping member 13 are combined, thereby realizing a configuration in which a plurality of protruding portions individually having an isosceles triangle shape are arranged (or a configuration in which a plurality of V-shaped groove portions are arranged). In addition, on the downstream side in the irradiating direction (Z-axis directional positive side), the bottom surfaces 13a of each damping member 13 are combined, thereby configuring a plane which spreads straight in the X-axis direction. Similarly, the side surface 13c of the damping member 13 in a case of being seen in the X-axis direction is bonded to the side surface 13c of another damping member 13. A cross-sectional shape in a case of being seen in the X-axis direction has the same conceptual configuration as a cross-sectional shape in a case of being seen in the Y-axis direction illustrated in FIG. 6A. As illustrated in FIG. 5A, in the main body portion 11 of the ridge filter 22, at a position different from that of the damping member 13 in a case of being seen in the irradiating direction, the pass-through portion 16 passing through the main body portion 11 in the irradiating direction is formed. The position different from that of the damping member 13 denotes a region in the main body portion 11 excluding a region where the damping member 13 is disposed. That is, in the main body portion 11, the region where the damping member 13 is not disposed corresponds to “the position different from that of the damping member 13”. In the present embodiment, the square pass-through portion 16 is formed at a part surrounded by edge portions 13e on the bottom surface 13a of each damping member 13. The pass-through portion 16 is formed such that diagonal lines are respectively parallel to the Y-axis direction and the X-axis direction. In addition, the plurality of pass-through portions 16 are arranged in the Y-axis direction at predetermined pitches and the plurality thereof are arranged in the X-axis direction at predetermined pitches. The edge portion 13e of the damping member 13 in a case of being seen in the XY-plane direction is separated from the edge portion 13e of another damping member 13 via the pass-through portion 16. In the present embodiment, as illustrated in FIG. 6B, the edge portion 13e of the damping member 13 in a case of being seen in a direction oblique 45° to the Y-axis directional negative side toward the X-axis directional positive side is separated from the edge portion 13e of another damping member 13 via the pass-through portion 16. On the upstream side in the irradiating direction (Z-axis directional negative side), the inclined surfaces 13b of each damping member 13 are combined, thereby realizing a configuration in which a plurality of protruding portions individually having an isosceles triangle shape are arranged. In addition, on the downstream side in the irradiating direction (Z-axis directional positive side), while the bottom surface 13a of each damping member 13 is divided by the pass-through portion 16, a plane which spreads straight in the X-axis direction is configured. Similarly, the side surface 13c of the damping member 13 in a case of being seen in the X-axis direction is bonded to the side surface 13c of another damping member 13. A cross-sectional shape in a case of being seen in a direction oblique 45° to the Y-axis directional positive side toward the X-axis directional positive side has the same conceptual configuration as a cross-sectional shape illustrated in FIG. 6B. The damping members 13 and the pass-through portions 16 are arranged so as to have the same shapes and to be in the same array in the entire region of the main body portion 11. In addition, the pattern structure formed by the damping members 13 and the pass-through portions 16 has symmetry with respect to the Y-axis and has symmetry with respect to the X-axis. Therefore, when being seen in the irradiating direction, in each of the regions in the main body portion 11, the ratio between an area occupied by the damping member 13 and an area occupied by the pass-through portion 16 becomes substantially uniform. Specifically, as illustrated in FIG. 4, a predetermined unit area SE is set with respect to the main body portion 11. Even in a case where the unit area SE is moved at random within the main body portion 11, the ratio between the area occupied by the damping member 13 and the area occupied by the pass-through portion 16 may be substantially uniform within the unit area SE. In a state where the ratio between the area occupied by the damping member 13 and the area occupied by the pass-through portion 16 is substantially uniform in such a manner, the strength of the main body portion 11 is ensured by causing the pyramid shapes to be interlinked with each other in a predetermined pattern in the Y-axis direction and the X-axis direction. Here, it is preferable that the main body portion 11 does not warp due to its own weight and the main body portion 11 maintains the strength to the extent that the posture of a flat plate shape can be maintained. Next, the operation and the effect of the charged particle beam therapy apparatus 1 and the ridge filter 22 according to the present embodiment will be described. First, with reference to FIGS. 7A and 7B, a ridge filter 100 according to a comparative example will be described. The ridge filter 100 according to the comparative example includes a plurality of damping members 111 which individually have the same cross-sectional shape and are arrayed in the X-axis direction, and a support member 112 which supports all the plurality of damping members 111. Each damping member 111 has a triangular cross-sectional shape protruding toward the upstream side in the irradiating direction of the charged particle beam R, and the cross-sectional shape extends along the Y-axis direction. In addition, as illustrated in FIG. 7B, a gap 113 is formed between the damping members 111 adjacent to each other. However, the support member 112 also spreads in a part corresponding to the gap 113. Therefore, the charged particle beam R incident on the ridge filter 100 passes through the support member 112 even in a place where the charged particle beam R does not pass through the damping member 111. Accordingly, there are cases where the charged particle beam R scatters due to the support member 112. In contrast, in the charged particle beam therapy apparatus 1 according to the present embodiment, the ridge filter 22 includes the plurality of the damping members 13 reducing the energy of the incident charged particle beam R, in the XY-plane direction which intersects the irradiating direction of the charged particle beam R. Here, the side surface 13c of the damping member 13 in a case of being seen in the XY-plane direction is bonded to the side surface 13c of another damping member 13. In this manner, the damping members 13 adjacent to each other support each other, so that even if there is provided no support member supporting all the damping members 13, it is possible to ensure the strength for serving as the ridge filter 22. Furthermore, since the strength can be ensured even if there is provided no support member, it is possible to form the pass-through portion 16 which passes through the ridge filter 22 in the irradiating direction at a position different from that of the damping member 13 in a case of being seen in the irradiating direction. According to such a structure, a charged particle beam R which is not incident on the damping member 13 can pass through the pass-through portion 16, and thus, the charged particle beam R can travel to the downstream side of the ridge filter 22 without scattering. Consequently, the charged particle beam R can be restrained from scattering. In the charged particle beam therapy apparatus 1 according to the present embodiment, the XY-plane direction has the X-axis direction and the Y-axis direction. The damping members 13 individually have a pyramid shape, are arranged along the X-axis direction, and are arranged along the Y-axis direction. According to such a configuration, even if there is provided no support member, it is possible to ensure the strength for serving as the ridge filter 22. In addition, since the damping members 13 individually have a pyramid shape and are arranged along the X-axis direction and the Y-axis direction, the damping members 13 are disposed in a two-dimensional array. For example, in the ridge filter 100 according to the comparative example as illustrated in FIGS. 7A and 7B, the damping members 111 extending straight in the Y-axis direction are arranged in the X-axis direction. Accordingly, there appears shade of a Bragg peak in the X-axis direction, and there appears no shade in the Y-axis direction, resulting in shade having a striped pattern in a case of being seen in a planar manner. Meanwhile, since the damping members 13 are disposed in a two-dimensional array, it is possible to obtain planar shade of the Bragg peak, so that the approximately even shade of the Bragg peak can be realized. In addition, the ridge filter 22 according to the present embodiment is the ridge filter 22 for a charged particle beam therapy apparatus generating a spread out Bragg peak of the charged particle beam R. The ridge filter 22 includes the plurality of damping members 13 reducing the energy of the incident charged particle beam R in the XY-plane direction which intersects the irradiating direction of the charged particle beam R. The damping member 13 has a cross-sectional area changing along the irradiating direction and has the side surface 13c in a case of being seen in the intersecting direction, being bonded to the side surface 13c of another damping member 13. The pass-through portion 16 passing through the ridge filter 22 in the irradiating direction is formed at a position different from that of the damping member 13 in a case of being seen in the irradiating direction. According to the ridge filter 22 of the present embodiment, it is possible to obtain an operation and an effect similar to those of the charged particle beam therapy apparatus 1. The embodiment of the invention is not limited to the embodiment. The shape and the array of the damping members are not limited to the embodiment. For example, as illustrated in FIGS. 8A and 8B, damping members 150 individually having a trigonal pyramid shape may be employed. In the example illustrated in FIG. 8A, the damping members 150 may be arranged in a state of being oriented toward a certain direction and may be arrayed in a pattern such that a triangular pass-through portion 156 is formed by edge portions 150e of three damping members 150. Alternatively, as illustrated in FIG. 8B, the damping members 150 may be arrayed in a pattern such that six damping members 150 are bonded to each other while varying the orientations and a hexagonal pass-through portion 157 is formed by the edge portion 150e of each damping member 150. In any of the examples in FIGS. 8A and 8B, the damping members 150 is bonded to an adjacent damping members 150 via a side surface formed in a part at a corner portion. In addition, as illustrated in FIG. 9A, a damping member 160 having a hexagonal pyramid shape is employed and is bonded to another damping member 160 via a side surface at a corner portion. A triangular pass-through portion 166 is formed by edge portions 160e of the damping members 160. As illustrated in FIG. 9B, a damping member 170 having an octagonal pyramid shape is employed and is bonded to another damping member 170 via a side surface. A square pass-through portion 176 is formed by edge portions 170c of the damping members 170. In addition, the damping member does not have to have a perfect pyramid shape as in the embodiment and modification examples. The damping member may be configured to have an approximated pyramid shape by combining stepped shapes. For example, a ridge filter illustrated in FIG. 10 includes damping members 180 configured to be approximated to the damping members 150 individually having a trigonal pyramid shape in the ridge filter illustrated in FIG. 8B. Each damping member 180 is configured to include a plurality of members 181 which individually have a substantially triangular plate shape and are stacked. The members 181 are gradually reduced in size toward the upstream side in the irradiating direction, while maintaining similar figures. Accordingly, the damping member 180 has a configuration in which the inclined surfaces of the trigonal pyramid are approximated with a plurality of stepped surfaces. Such a damping member 180 also corresponds to the shape of which the cross-sectional area changes along the irradiating direction. In all the cases of FIGS. 8A, 8B, 9A, and 9B, the damping members are arrayed in a particular pattern along the X-axis direction and the Y-axis direction. Therefore, in each region of the ridge filter, the ratio between the area occupied by the damping member and the area occupied by the pass-through portion can be substantially uniform. In a case other than those of FIGS. 8A, 8B, 9A, and 9B, damping members having any shape and array pattern may be employed. The damping members of each the ridge filter form a geometric pattern which satisfies plane-filling of Archimedes. In addition, all the pyramid shapes of the damping members in the above-described examples may be replaced by cone shapes. It should be understood that the invention is not limited to the above-described embodiment, but may be modified into various forms on the basis of the spirit of the invention. Additionally, the modifications are included in the scope of the invention. |
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summary | ||
abstract | The present invention relates to a device for phase stepping in phase contrast image acquisition, the device (1) comprising: a mobile grating (10); a guiding element (11); a restoring element (12); and a locking element (13); wherein the guiding element (11) is configured to guide the mobile grating (10) between a first position (2) and a second position (3); wherein the restoring element (12) is configured to apply a force to the mobile grating (10); wherein the force is directed from the first position (2) to the second position (3); and wherein the locking element (13) is configured to releasably lock the mobile grating (10) in the first position (2). In an example, during the motion of the mobile grating (10) back to equilibrium, a decoder (11a) for the position of the mobile grating (10) along the guiding element (11) may trigger at least four measurement frames over a period of at least 2*Pi. The invention provides a device (1) for phase stepping in phase contrast image acquisition which provides a fast image acquisition without a significant delay and which reduces positional inaccuracies and which avoids back-lash. |
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040381366 | summary | This invention relates to a structure for supporting the lateral neutron shield system of a fast reactor core. It is known that the core of a fast reactor which is cooled with liquid sodium is surrounded by a neutron shield system. The top and bottom shield are formed by elements which are introduced directly above and beneath the fuel assemblies which constitute the reactor core. The lateral shield of the reactor core is constituted by elements which have the same shape as the fuel assemblies and surround the reactor core. A better understanding of the problem to be solved will be obtained by referring to FIG. 1 of the accompanying drawings, in which is shown diagrammatically the lower portion of a sodium-cooled fast reactor. There can be seen in this figure the lower portion 2 of the metallic vessel which contains the complete reactor assembly, that is to say essentially the reactor core 4, the primary sodium and the primary pumps 5 and the primary heat exchangers 3. The reactor core 4 rests on a diagrid 6 in which the bottom end-fittings of the fuel elements constituting the core 4 are inserted. The diagrid also permits the flow of cold primary sodium within said fuel assemblies, said primary sodium being injected into the diagrid 6 through ducts such as the duct 8 which are connected to the outlets of the primary pumps 5. The diagrid rests on a support grid 10 which is rigidly fixed to the wall 2 of the reactor vessel by means of a frusto-conical shell-plate 12. The lateral neutron shield is generally designated by the reference numeral 14 and surrounds the reactor core. The neutron shield elements are, for example, round stainless steel members of hollow construction, the top portion of which is level with the top portion of the fuel elements constituting the reactor core 4. The bottom end-connectors of the lateral shield elements are inserted in a so-called "false grid" 16 or ring-shaped support structure which rests on the periphery of the diagrid 6. The false grid is higher than the diagrid and surrounds the reactor core 4, with the result that a peripheral side restraint is applied to the base of the core. The design function of the false grid 16 is therefore to ensure that the neutron-shielding elements 14 are rigidly maintained in position and that said elements can also be cooled by the liquid sodium. In addition, the false grid must afford peripheral side restraint for the reactor core. One known false grid arrangement consists in designing this latter in six sectors which are not joined together but connected to the diagrid by means of a system of keys. Each sector is formed by a bottom plate and a top plate which are pierced by bores at the bottom of the system of shield assemblies and separated by hollow spacer members placed at the level of said bores. The bottom end-connectors of the shield assemblies are intended to be fitted in the bores of said spacer members. Tie-rods which serve to connect the top and bottom plates ensure powerful clamping of these latter against the spacer members. Sodium coolant is supplied to the lateral shield assemblies through lateral holes which are formed in said spacer members and which must therefore occupy precise positions both angularly and in height. For this reason, after the spacer members have been machined to the required tolerances, they must accordingly be oriented on the false grid at the time of assembly. This result is relatively difficult to obtain and calls for the presence of positioning studs carried on the top face of the bottom plate. Moreover, by reason of the fact that the sectors are independent, the false grid does not permit of effective application of peripheral side restraint to the fuel assemblies of the reactor core. The precise aim of the present invention is to provide a support structure for the lateral shield system of a fast reactor core which overcomes the disadvantages mentioned in the foregoing by making it possible in particular to provide a peripheral side restraint for the reactor core, simple forms of construction and higher structural rigidity of the core as a whole and of its neutron shield systems. The structure for supporting the lateral neutron shield system of a fast reactor core is distinguished by the fact that the structure has the shape of a horizontal hexagonal ring whose internal contour coincides exactly with the external contour of the reactor core, the bottom face of said ring structure being intended to rest on the periphery of the core support structure, said ring structure being constituted by a plurality of layers of metallic plates, said layers being maintained in relative positional relation by means of clamping members, the metallic plates of one layer being angularly displaced with respect to the plates of adjacent layers, said ring structure being penetrated by vertical through-holes in which are fitted the bottom end-connectors of the elements constituting the lateral shield system. In a first embodiment, the layers of metallic plates are contiguous. In a second embodiment, two successive layers of metallic plates are separated by washers of small thickness. As a preferable feature, the members for clamping the layers together consist of tie-bolts. |
abstract | The present invention is generally directed to various reticle writing methodologies to reduce write time, and a system for performing same. In one illustrative embodiment, the method comprises exposing a layer of photoresist in accordance with a first writing pattern in a first area of the layer of photoresist and exposing the layer of photoresist in accordance with a second writing pattern in a second area of the layer of photoresist, the first and second areas of the layer of photoresist overlapping one another in at least one region. In another illustrative embodiment, the method comprises creating a collection of digital data corresponding to a desired pattern for a reticle and separating the collection of digital data into at least two separate groups of data, a first of the data groups being used to define a first writing pattern for the reticle, a second of the data groups being used to define a second writing pattern for the reticle, wherein the first and second writing patterns overlap one another in at least one region. In yet another illustrative embodiment, the method comprises forming a layer of photoresist above at least one of a semiconducting substrate and a process layer, exposing the layer of photoresist in accordance with a first writing pattern in a first area of the layer of photoresist, and exposing the layer of photoresist in accordance with a second writing pattern in a second area of the layer of photoresist, wherein the first and second areas overlap one another in at least one region. |
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claims | 1. A filter for separating particles from cooling water in a nuclear plant of a light water type,wherein the filter has an inlet end and an outlet end and is arranged to permit through-flow of the cooling water in a main flow direction from the inlet end to the outlet end,wherein the filter includes a number of sheets, which extend in the flow direction from the inlet end to the outlet end,wherein said sheets are arranged beside each other and form passages for the cooling water through the filter front the inlet end to the outlet end, andwherein said sheets include a first portion, which extends from the inlet end, a second portion, which extends from the outlet end, and a third portion, which extends between the first portion and the second portion,wherein said sheets along the first portion have a wave-shape formed by a number of successive waves and extending in a transverse direction transversally to the flow direction, andwherein said sheets along the third portion have a wave-shape formed by a number of successive waves and extending in the flow direction,characterized in that each wave of said wave-shape of the first portion has a maximum amplitude, wherein the maximum amplitude decreases continuously in the flow direction towards the third portion, and that said maximum amplitude is zero at a transition to the third portion. 2. A filter according to claim 1, characterized in that said wave-shapes are continuous. 3. A filter according to claim 1, characterized in that said sheets along the second portion have a wave-shape formed by a number of waves and extending in said transverse direction transversally to the flow direction. 4. A filter according to claim 1, characterized in that said sheets along the first portion ore arranged beside each other in such a way that each pair of adjacent sheets abuts each other at valleys and ridges, respectively, of said wave-shape, wherein each passage between two adjacent sheets forms a plurality of inlet channels arranged beside each other. 5. A filter according to claim 4, characterized in that said sheets along the second portion are arranged beside each other in such a way that each pair of adjacent sheets abuts each other at valleys and ridges, respectively, of said wave-shape, wherein each passage between two adjacent sheets forms a plurality of outlet channels arranged beside each other. 6. A filter according to claim 1, characterized in that said sheets are connected to each other at at least one point at said valleys and ridges, respectively. 7. A filter according to claim 6, characterized in tat said sheets are connected to each other by means of a fuse weld where the sheets abut each other. 8. A filter according to claim 6, characterized in that said sheets are connected to each other by means of a spot weld where the sheets abut each other. 9. A filter according to claim 3, characterized in that each wave of said wave-shape of the second portion has a maximum amplitude, wherein the maximum amplitude decreases continuously in the direction towards the third portion. 10. A filter according to claim 5, characterized in that each inlet channel has substantially the same flow area as each outlet channel. 11. A filter according to claim 5, characterized in that the center line of each inlet channel is concentric with the center line of a respective corresponding outlet channel. 12. A filter according to claim 1, characterized in that the third portion forms an intermediate channel between two adjacent sheets. 13. A filter according to claim 12, characterized in that the third portion includes projections extending into the intermediate channel. 14. A filter according to claim 11, characterized in that said center line of the inlet channel and the outlet channel extends between two adjacent projections of the third portion. 15. A filter according to claim 13, characterized in that said projections are firmed through plastic deformation of the sheet. 16. A filter according to claim 13, characterized in that said projections include a tab, which is out from the sheet. |
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039390390 | claims | 1. For use in a nuclear reactor having a core, a clamping device for clamping a plurality of core elements together in forming the core, comprising: projection means being movably mounted on a core element for movement between a withdrawn position and a position projecting outwardly from a core element toward another core element when the core elements are arranged within the core of a nuclear reactor; operating means for moving said projection means from its withdrawn position to its projected position a predetermined length to clamp the core element in cooperation with other core elements of the nuclear reactor core; a wrapper tube forming the exterior of the core element mounting with its longitudinal axis extending generally vertical in the reactor core and being of hexagonal shape in horizontal cross section; said projection means being provided adjacent the upper portion of said wrapper tube; said operating means being mounted within said wrapper tube for vertical movement; said wrapper tube having a hole within one of its side walls for said projection means; said projection means moving generally horizontally from within said wrapper tube in its withdrawn position horizontally through its hole to project horizontally beyond said hole and outside of said wrapper tube in its projected position; and separate lock means for locking said vertically moving operating means and said wrapper tube against relative movement in one of said projection means positions. 2. The apparatus of claim 1, wherein a separate one of said projection means is provided in each of the side walls of said hexagonal wrapper tube. 3. The apparatus of claim 2, wherein said operating means simultaneously moves all of said projection means. 4. The apparatus of claim 2, wherein a plurality of core elements are arranged with their axes parallel with respect to each other in the core of the nuclear reactor with six core elements surrounding a clamping core element respectively adjacent the six walls of the hexagonal clamping core element, further wherein only the central clamping core element of the seven core elements is provided with said projection means and said operating means, and further wherein said pattern of seven core elements is repeated throughout the reactor core. 5. The apparatus of claim 1, wherein a first set of said projection means respectively for all of said hexagonal wrapper tube side walls is provided generally adjacent the upper end of said wrapper tube, and further including a second set of said projection means for all of said hexagonal wrapper tube side walls being provided generally adjacent the middle portion of said wrapper tube. 6. The apparatus of claim 5, wherein said operating means simultaneously moves all of said projection means. 7. For use in a nuclear reactor having a core, a clamping device for clamping a plurality of core elements together in forming the core, comprising: projection means being movably mounted on a core element for movement between a withdrawn position and a position projecting outwardly from the core element toward another core element when the core elements are arranged within the core of a nuclear reactor; operating means for moving said projection means from its withdrawn position to its projected position a predetermined length to clamp the core element in cooperation with other core elements of the nuclear reactor core; a wrapper tube having a longitudinal axis to be parallel with the longitudinal axes of other wrapper tubes when mounted within the nuclear reactor core and having a hexagonal shape in cross section transverse to said longitudinal axis; said projection means including a plurality of pawls pivotally connected to said wrapper tube, and each pawl having a configured cam slot spaced from said pivotal connection; said operating means including an operating tube mounted within said wrapper tube for axial longitudinal movement with respect to said wrapper tube, and said operating tube having a plurality of arms mounting a plurality of pins each in engagement within a corresponding curved configured cam slot of said pawls; said wrapper tube having a hole in its side walls immediately adjacent each of said pawls; and said operating means projecting said pawls through their respective wrapper tube holes by movement of said operating tube axially with respect to said wrapper tube to correspondingly move said pins in the same direction relative to said cam slots for camming said pawls about their pivotal connections through their respective holes. 8. The apparatus of claim 7, including a plurality of guide blocks fixedly mounted on the interior of said wrapper tube for guidingly engaging said operating tube with axial sliding contact; each of said plurality of pawls being pivotally connected respectively to each of said guide blocks. 9. The apparatus of claim 7, wherein the upper portion of said operating tube is provided with operating head means for engagement with a gripper to move said operating tube axially with respect to said wrapper tube. 10. The apparatus of claim 7, further including lock means for selectively locking said operating tube and said wrapper tube against axial movement in one of the positions of said projection means, said lock means comprising a plurality of rods axially mounted within bores in the upper portion of said operating tube, first spring means for biasing said rods axially in one direction, a lock operating plate drivingly connected to said rods for moving said rods axially in the opposite axial direction against the spring bias, a plurality of lock pins slidably mounted within bores of said operating tube for movement transversely with respect to the movement of said rods, a plurality of apertures in the side wall of said wrapper tube being aligned with said lock pins in said one position of said projection means, second spring means for biasing said lock pins transversely into engagement with said apertures, respectively, and means provided on each of said rods and each of said lock pins for converting the axial movement of each of said rods into the transverse movement of said lock pins so that movement of said rod operating plate in said opposite direction will correspondingly move all of said rods in their direction against their spring force, to correspondingly move all of said lock pins against their spring bias in the direction to withdraw them from the apertures in said wrapper tube, respectively. 11. The apparatus of claim 10, including a plurality of guide blocks fixedly mounted on the interior of said wrapper tube for guidingly engaging said operating tube with axial sliding contact; each of said plurality of pawls being pivotally connected respectively to each of said guide blocks. 12. The apparatus of claim 7, wherein said wrapper tube is hexagonal; a projection means pawl is provided for each of the side walls of said hexagonal wrapper tube for respectively extending therethrough adjacent one longitudinal end. 13. The apparatus of claim 12, including a second set of pawls generally identical to said previously mentioned pawls and provided for each of the wrapper tube side walls generally adjacent the middle portion of said wrapper tube. 14. The apparatus of claim 12, wherein a plurality of said core elements are mounted with their wrapper tubes and operating tubes generally axially parallel with respect to each other, and further wherein said pawls project transversely from their respective wrapper tube side walls approximately one-half the gap distance between the associated adjacent wrapper tube walls. |
043671950 | summary | BACKGROUND OF THE INVENTION The invention relates to an apparatus for circumferentially homogenizing temperatures of a ferrule of a component traversing the upper slab of a nuclear reactor. More specifically, the present invention relates to an apparatus of this type applied to the case of components (circulating pumps, intermediate exchanger) mounted in a fast nuclear reactor cooled by a liquid metal, said reactor being of the integrated type. In order to provide a better understanding of the problem solved by the present invention, reference is advantageously made to the attached FIGS. 1 and 2, which respectively show a fast neutron reactor of the integrated type in vertical section and a detailed view showing how a pump or intermediate exchanger traverses the slab. FIG. 1 shows in simplified form the main reactor vessel 2 suspended on the upper concrete slab 4 provided with its system or rotary plugs 6. The main vessel 2 contains the inner vessel 8, which in turn contains the core 10 and the hot liquid metal (e.g. liquid sodium) leaving the core, the liquid metal level being designated by N. Above level N and below slab 4 there is an inert covering gas cushion 11 the gas being for example argon. The hot liquid metal enters intermediate exchangers such as 12, which are suspended on slab 4 and which traverse the latter by not shown cylindrical passages. In a similar way, circulating pumps such as 14 are suspended on slab 4 and traverse the latter by cylindrical passages 16. The invention relates to the problems linked with the traversal of the said slab. FIG. 2 shows in a more detailed manner the passage of a pump 14 through slab 4. Pump 14 is surrounded by a pump ferrule 14a and slab 4, more particularly level with passage 16 is covered by a sheet 18. The slab is cooled by water pipes 20. Between pump ferrule 14a and the covering sheath 18, there is an intermediate ferrule 22 defining an outer annular space 24 and an inner annular space 26 (the same applies for the passage of an intermediate exchanger). The outer space 24 is insulated from the gaseous cover 11 by a hydraulic joint system (liquid sodium) with a container 28. Thus, there are no problems for the outer space 24, which is insulated from the remainder of the gaseous mass. Inner space 26 communicates with the gaseous cover 11. Inner space 26 contains "open" thermosiphons, which are supplied by the inert gas in cover 11. These thermosiphons have an upward flow of hot gas and a downward flow of cool gas. Their characteristics are dependent on parameters which are at present not well known. In the ferrule of the component, (intermediate exchanger or pump) these thermosiphons produce thermal indentations, which can be displaced in a circular manner (in a horizontal plane) and which thus create thermal cycles prejudicial to the good mechanical behaviour of the corresponding ferrules. It is therefore necessary to eliminate or at least reduce these circumferential thermal gradients. For example, in the case of the French Super Phenix reactor, this circumferential gradient on the pump ferrule is estimated to be 200.degree. C. One solution would involve choosing all the ferrules with a material having a high thermal conductivity. However, such a solution is unacceptable. On the one hand, this is because the choice of such a material is considerably limited for cost and mechanical behaviour reasons and on the other and in particular because it is necessary to maintain a high vertical thermal gradient between the space 11 or the roof of the pile and the top of the slab 4. Thus, the upper end of the slab must be kept at a maximum temperature of approximately 50.degree. C. BRIEF SUMMARY OF THE INVENTION The present invention specifically relates to an apparatus making it possible to homogenize the temperature of the ferrule of the component, i.e. reduce the horizontal thermal gradient in said ferrule, whilst maintaining the necessary vertical thermal gradient, whilst utilizing conventional nuclear material for the construction of the ferrules. The apparatus essentially comprises at least one assembly forming a horizontal or substantially horizontal heat pipe located within the ferrule and in thermal contact with the latter. Preferably, there are numerous heat pipes spaced over the height of the ferrule. It is possible either to use one and the same heat pipe positioned over substantially the entire circumference of the ferrule or a plurality of heat pipes placed end to end, whereby each heat pipe covers a fraction of the circumference. In the case where the heat pipes are horizontal, each assembly forming a heat pipe constitutes a ring and thus defines a ferrule temperature homogenization level. In the case where the heat pipes are slightly inclined (e.g. by approximately 2.degree.), the heat pipe systems form a helix with a plurality of threads, each thread being formed by a number of heat pipes arranged end to end. If each heat pipe system comprises a plurality of heat pipes arranged end to end and when each heat pipe only covers a portion of the circumference, it can be advantageous to superimpose two rings of heat pipes in such a way that the ends of the heat pipes of one ring are displaced relative to the ends of the heat pipes of the other ring. |
abstract | A boiling water reactor has a core disposed in the reactor pressure vessel and loaded with a plurality of fuel assemblies including transuranic nuclides. A ratio of Pu-239 in all of the transuranic nuclides included in the fuel assembly, which is loaded in the core, with a burnup of 0 is 3% or more but 45% or less. In the fuel assembly having a channel box and a plurality of fuel rods disposed in the channel box, a transverse cross section of a fuel pellet in the fuel rod occupies 30% or more but 55% or less of a transverse cross section of a unit fuel rod lattice in the channel box. |
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050193210 | claims | 1. A fusion power generating means for generating thermal energy from fusion reactions in an ionized plasma of fusible fuel comprising: (a) a fusion core unit including as components thereof, (b) disconnectable means extending into said fusion core unit for delivering said fusible fuel therein; (c) disconnectable means for transporting a cooling fluid to and through the toroidal field coils of said fusion core unit; (d) a power supply connected through disconnectable means to said toroidal field coils for generating a toroidal magnetic field for confining the ionized plasma within the plasma cavity of said fusion core unit, said toroidal magnetic field having a strength of on the order of greater than 100 KG; (e) said plurality of toroidal field coils comprising high-strength, non-superconducting conductors for sustaining said toroidal magnetic field and for withstanding said thermal energy; (f) blanket means, positioned completely outside of and substantially surrounding said toroidal field coils, said blanket means comprising at least two modules; (g) said toroidal field coils and said plasma cavity have no blanket means therebetween; (h) means for connecting and disconnecting each of said disconnectable means, means for separating said at least two modules of said blanket means a distance sufficient to allow removal of the fusion core unit, means for removing and inserting a fusion core unit as a single entity between the separated modules of said blanket means; (i) blanket cooling fluid transport means connected to said blanket means for transporting a cooling fluid to and through said blanket means; and (j) means, connected to at least one of the blanket cooling fluid transport means and the toroidal field cooling fluid transport means, for extracting thermal energy therefrom. 2. A fusion power generating means as recited in claim 1, wherein said fusion core unit includes a toroidal housing having a major radius on the order of 50 cm and a minor radius on the order of 20 cm. 3. A fusion power generating means as recited in claim 1, wherein said plasma cavity further includes a shell surrounded by said toroidal field coils and said means for transporting a cooling fluid to and through said plurality of toroidal field coils further comprises means for transporting said cooling fluid around surfaces of said toroidal shell adjacent said toroidal field coils. 4. A fusion power generating means as recited in claim 1, wherein said blanket means further include tritium breeding means for generating tritium from neutrons emitted by said fusion reactions. 5. A fusion power generating means as recited in claim 1, wherein said plasma cavity further includes a shell surrounded by said toroidal field coils. 6. A fusion power generating means as recited in claim 5, including means in addition to said ohmic heating means for generating said poloidal magnetic field. 7. A fusion power generating means as recited in claim 5, wherein said fusion core unit comprises a toroidal region having an aspect ratio of about on the order of 2.5. 8. A fusion power generating means as recited in claim 5, wherein said plurality of toroidal field coils comprise copper coils. 9. A fusion power generating device as recited in claim 1, wherein said toroidal magnetic field strength is on the order of 100-150 kg. 10. A fusion power generating device as recited in claim 1, wherein said means for extracting thermal energy is connected to both said blanket cooling fluid transport means and said toroidal field coil cooling fluid transport means. 11. A fusion power generating device as recited in claim 10, wherein said thermal energy extraction means comprises a first fluid transport means for extracting thermal energy from said toroidal field coils and a second fluid transport means for extracting thermal energy from said blanket means. 12. A fusion power generating device as recited in claim 10, wherein said thermal energy extraction means comprises a single cooling fluid transport means for the toroidal field coils and for said blanket means. 13. A fusion power generating device as recited in claim 12, wherein said single fluid transport means connects the cooling fluid transport means for the toroidal field coils and the cooling fluid transport means for the blanket means in series with one another and with a heat exchange means. 14. A fusion power generating device as recited in claim 12, wherein said single fluid transport means separately connects the cooling fluid transport means for the plurality of toroidal field coils and the cooling fluid transport means for the blanket means to a heat exchange means. |
claims | 1. A laboratory device unit for at least one of processing or analyzing substances, mixtures, or media, comprising a laboratory device (3) having at least one of a scale, a heater, or an agitator, with at least first and second different remote controls (2, 6) configured to control existing or adjustable parameters or detectable measurements in the laboratory device (3) including for at least one of the scale, the heater, or the agitator, the laboratory device (3) the first and second different remote controls are configured to control at least one of a capacity, a performance spectrum, a range of adjustments, or a range of measurements with regards to the adjustable parameters or detectable measurements for the at least one of the scale, the heater, or the agitator, and the first remote control (2) for operating the laboratory device (3) includes controls that are configured to allow operation of the at least one of the scale, the heater, or the agitator with respect to at least one of the capacity, the performance spectrum, the range of adjustments, or the range of measurements and the second remote control (6) for operating the laboratory device (3) includes controls that are configured to allow operation of the at least one of the scale, the heater, or the agitator with a reduction to at least one of the capacity, the spectrum, the range of adjustments, or the range of measurements. 2. A laboratory device unit according to claim 1, wherein the first remote control (2) is configured to operate at at least one of a full capacity, a full performance spectrum, a full range of adjustments, or a full range of measurements of the laboratory device (3). 3. A laboratory device unit according to claim 1, wherein at least one additional remote control (6) is provided, which is configures to operate at interim values with regards to at least one of the capacity, the performance spectrum, the range of adjustments, or the range of measurements of the laboratory device (3). 4. A laboratory device unit according to claim 1, wherein an additional remote control (6) is provided, by which only a portion of the parameters or measurements of the laboratory device (3) are accessed or addressed or controlled. 5. A laboratory device according to claim 1, wherein all of the remote controls (2, 6) are configures to operate the full range of the existing or adjustable parameters or detectable measurements and the second remote control (2, 6) provided for recalling only a portion of the capacities or the range of the adjustment or the range of measurements, includes electronic or software blocking of higher valued functions. 6. A laboratory device unit according to claim 5, wherein the second remote control (2, 6) with higher valued functions blocked by electronics or software includes a feature for release of the electronic or software blocking. 7. The laboratory device unit according to claim 1, wherein the remote controls configured to operate the laboratory device (3) having different capacities or different performance spectrums or different ranges of adjustments or different ranges of measurements comprise different displays (4), with the second remote control (2, 6) for the lower capacity or the lower performance spectrum or the smaller ranges of adjustments or the smaller ranges of measurements comprising a simple display (4) and the first remote control (2, 6) for the higher performances and measurement ranges comprising a more expensive display (4). 8. A laboratory device unit according to claim 7, wherein the displays on the remote controls (2, 6) are interchangeable. 9. A laboratory device unit according to claim 1, wherein the remote control (2, 6) allocated to the laboratory device (3) are fixed thereon in a detachable fashion. 10. A laboratory device unit according to claim 9, wherein a fixation for the remote control (2, 6) to the laboratory device (3) includes contacts to charge a battery serving as a power supply for the remote control (2, 6). 11. A laboratory device unit according to claim 10, wherein the fixation comprises contacts for the laboratory device (3) at which the remote control (2, 6) is fixed to address the remote control (2, 6). 12. A laboratory device unit according to claim 9, wherein in a holding position, the remote control (2, 6) respectively fixed to the laboratory device (3) in a detachable fashion is at least one of mechanically or electrically coupled to components of the laboratory device (3) creating the parameters or the measurements. 13. A laboratory device unit according to claim 12, wherein a radio connection between the laboratory device (3) and the remote control (2, 6) is deactivated for the mechanically or electrically coupled remote controls (2, 6) and operating signals are transmitted via plug-in connections. 14. A laboratory device unit according to claim 1, wherein the laboratory device comprises a weighing function as a parameter, that is implemented by support feet located at a bottom and connected or allocated to a weighing device. 15. A laboratory device unit according to claim 1, wherein the laboratory device is a magnetic agitator (3), which is controllable with the second remote control (6) of lower capacity or lower performance spectrum at a lower rotation or within a range of lower rotations or lower heating temperature than values possible for the magnetic agitator (3) and that maximum rotations or heating temperatures or a weighing function are controllable or accessible by at least the first remote control (2). 16. A laboratory device unit according to claim 1, wherein certain control processes with changing parameters or measurements are performed using the remote control (2) for higher capacities. 17. A laboratory device unit according to claim 1, wherein the laboratory device is an agitator or a dispersing device (3) and a maximum rotation or temperature is adjustable with the first remote control (2) and with at least the second remote control (6) only values are adjustable for the rotation or temperatures that are reduced in reference to maximum values. 18. A laboratory device unit for at least one of processing or analyzing substances, mixtures, or media, comprising a laboratory device (3), with a remote control (2, 6) configured to control existing or adjustable parameters or detectable measurements in the laboratory device (3) including for at least one of a scale, a heater, or an agitator, the laboratory device (3) is provided with at least one of a capacity, a performance spectrum, a range of adjustments, or a range of measurements with regards to the adjustable parameters or detectable measurements for the at least one of the scale, the heater, or the agitator, and the remote control (2) for operating the laboratory device (3) is configured to operate a full range of the existing or adjustable parameters or detectable measurements in the laboratory device, and the remote control includes electronic or software configured to block certain functions to allow a limited range of operation of the laboratory device with respect to the adjustable parameters or detectable measurements. |
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041526023 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to restraint devices and more particularly provides a flexible lateral restraint device for an underwater nuclear fuel storage rack. 2. Description of the Prior Art Nuclear power generating plants are typically fueled by elongated fuel assemblies, such as those including a bundle of nuclear fuel rods. Subsequent to utilization of the fuel assemblies in a nuclear core they are stored within fuel racks positioned within an enclosure such as a spent fuel pit. A typical spent fuel pit or pool includes leak tight vertical walls and a floor made of concrete and other support materials lined with stainless steel. The fuel racks include a rectangular, closely spaced array of cells, each cell sized to receive a fuel assembly. Because the assemblies have been irradiated they must be shielded, and water is typically utilized for this purpose. The assemblies and fuel racks are therefore maintained submerged within water within the pit. It is of significant importance that the fuel racks be restrained laterally within the fuel pit under accident conditions, such as a large seismic occurrence. Excessive deflections of the racks and contained fuel assemblies could result in damage to the assemblies and the surrounding environs. It is also recognized that differential thermal expansions can occur in time between the fuel racks and the walls of the containing fuel pit. Accordingly, several types of lateral restraint devices have been proposed in the past, which respond to these differential expansions as well as seismic occurrences. One type has been proposed which is affixed to the side of the fuel rack and positions, through an arc-like motion between the fuel rack and the containing wall, a load pad surface against the fuel pit wall. A pre-load is imposed upon the pads by a series of Belleville-type springs between a portion of the restraining device and the load pads. With such arc-type devices the pre-load is very difficult to adjust and the device is difficult to properly and accurately position. An excessive amount of space between the fuel rack and the pit wall is also required to accommodate the arc-like motion. Further, the arc motion devices are difficult to properly seat against the pit wall due to the high frictional forces and slippage among the contact pad surface and the pit wall. A significant improvement upon the arc motion device utilizes the substantially horizontal motion of a device affixed to the side of the rack and remotely extendible into contact with the pit wall. It operates in a manner similar to a scissors jack. This device is described in detail in application Ser. No. 789,912, filed Apr. 22, 1977, in the name of June S. Knight, Personal representative of the estate of Charles B. Knight, deceased, and entitled "Nuclear Fuel Rack Lateral Support and Pre-Load Device". Among the prior art, the arc-motion devices imposed loads between the racks and walls resulting from differential expansions, as well as imposing loads for pre-load purposes. Thus, the pit walls, for example, are loaded needlessly high. The pre-loaded scissors jack device suffers the same deficiency. Although these prior art devices will perform the intended function, it is desirable to have an alternative and simple lateral support which avoids unnecessarily high loadings, particularly in new plant installations where remote positioning capability is not necessarily required. SUMMARY OF THE INVENTION This invention provides a lateral support for nuclear fuel racks positioned underwater which provides the capability of a substantially rigid support in the event of large accidental loadings, while also providing the capability to inherently absorb differential thermal expansions between a fuel rack and the adjacent enclosing wall. The device can be manufactured simply, and can be sized to require a small amount of space. The device includes, similar to prior art systems, one or more load pads with a flat surface that can be positioned into contact with a vertical surface such as the wall of a spent fuel pit. The structure utilized for extending and positioning the pads, however, is significantly different and more simple than prior art structures and can eliminate the use of Belleville-type springs in a preferred embodiment. The pads are joined directly or preferably through a leveling foot to a sliding piston. The piston is disposed partially or completely within a horizontally oriented cylinder affixed to the side of the spent fuel rack. At the opposite end of the piston, an elastic structure such as a compression spring is positioned so as to continuously apply a positioning force to the piston. The device also includes structure for maintaining the piston in a fixed position within the cylinder with the spring in a significant compressed orientation. This structure can include a simple pin inserted through an opening in the wall of the cylinder matingly engaged with a receiving aperture in the piston. Means are also provided for controlling the amount of fluid, such as the water in which the assemblies and racks are immersed, which can flow into or out of the cylinder portion behind the piston in a predetermined fashion. This can be accomplished by contouring the piston to the shape of the cylinder, with a preselected clearance between the two components, or by incorporating selectively sized flow relief openings. The properly sized lateral support device can therefore be installed onto a fuel rack with the spring compressed and the piston maintained in place by the retaining pin. Subsequently, the holding pin is removed, allowing the spring to laterally move the piston and affixed surface so that the surface contacts the pit wall. Differential thermal expansions between the pit wall and the side of the fuel rack will then be accommodated by movement of the piston and the respective expansion or contraction of the compression spring. The water in the pit will also flow into the area behind the piston within the cylinder containing the compression spring. Under a large lateral loading, such as a seismic accident condition, the substantially incompressible fluid within the cylinder will resist and damp the motion of the piston, thereby providing a substantially rigid restraint between the fuel rack and the wall so as to alleviate the potential for excessive loading or bending forces upon the contained fuel assemblies. In view of the prior art, it is readily apparent that forces imposed between the fuel racks and pit walls from differential thermal expansion are avoided by water bleeding out of the cylinder. However, for a seismic event the substantially incompressible liquid will act as a rigid restraint. |
claims | 1. A method of scanning a substrate through an ion beam in an ion implanter, comprising:causing relative motion between the substrate and the ion beam; androtating the substrate substantially about its centre while causing the relative motion, such that the ion beam would pass over all of the substrate even if the substrate were not rotating, the relative motion between the substrate and the ion beam being at a constant speed. 2. The method of claim 1, comprising rotating the substrate substantially about its centre at a constant angular velocity. 3. The method of claim 1, wherein the ion beam is a ribbon beam, and the method comprises causing a relative motion between the substrate and the ribbon beam such that all of the substrate passes through the ribbon beam in a single pass. 4. The method of claim 3, comprising rotating the substrate such that it performs at least a complete revolution during the pass. 5. The method of claim 4, comprising rotating the substrate such that it performs at least fifteen to twenty revolutions during the pass. 6. The method of claim 1, wherein the ion beam is a spot ion beam, the method comprising causing the relative motion between the substrate and the ion beam such that the ion beam passes over all of the substrate by causing a series of translations of the substrate relative to the ion beam such that the ion beam traces a series of scan lines over the substrate. 7. The method of claim 6, comprising causing the translations to form the series of scan lines to be parallel or substantially parallel. 8. The method of claim 7, comprising causing the translations to form a raster pattern or saw-tooth pattern of scan lines. 9. The method of claim 7, comprising causing the translations to form a series of overlapping scan lines, such that the edges of the path of the ion beam across the substrate at the start of adjacent scan lines overlap. 10. The method of claim 7, comprising forming the scan lines to be spaced with a pitch P and wherein the ion beam has a dimension D in the direction of the pitch spacing, such that P is greater than 0.9 D. 11. The method of claim 7, comprising forming the scan lines to be spaced with a pitch P and wherein the ion beam has a dimension D in the direction of the pitch spacing, such that D is substantially equal to P. 12. The method of claim 6, comprising rotating the substrate such that it performs at least a complete revolution as the ion beam scans across each scan line. 13. The method of claim 12, comprising rotating the substrate such that it performs at least fifteen to twenty revolutions as the ion beam scans across each scan line. 14. The method of claim 12, comprising rotating the substrate and/or causing the relative motion between the substrate and the ion beam such that the resulting spirals traced by the ion beam over the substrate overlap on adjacent revolutions. 15. The method of claim 1, comprising causing relative motion between the substrate and the ion beam to form a scan line such that a point in the ion beam, having an average value of the total ion beam current along a section taken through the ion beam orthogonal to the direction of relative motion, passes over the centre of the substrate. 16. An ion implanter controller arranged to perform the method of claim 1. 17. An ion implanter comprising the controller of claim 16. 18. A computer program stored on a computer-readable medium and comprising computer program instructions that, when executed by a computer in an ion implanter, cause the ion implanter to perform the method of claim 1. 19. A computer readable medium carrying thereon a computer program according to claim 18. |
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abstract | A system that diagnoses a failure in a computer system is described. During operation, the system tests the computer system using a sequence of tests, where a given test includes a given load associated with a pre-determined failure mechanism for a given failure condition. During the given test, the system obtains results, which include telemetry signals that are monitored within the computer system. If the results indicate the given failure condition, the system ceases the testing and indicates that the computer system has the given failure condition. Otherwise, the system continues the sequence of tests until the sequence is completed, at which point, if no fault has been detected, the system indicates that a no-trouble-found (NTF) condition exists. |
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description | This application claims the priority, under 35 U.S.C. §119, of German patent application DE 10 2011 088 429.7, filed Dec. 13, 2011 and German patent application DE 10 2012 205 013.2, filed Mar. 28, 2012; the prior applications are herewith incorporated by reference in their entirety. Field of the Invention The invention relates to a device for repairing a damaged area in an underwater wall region of a container or tank, in particular in the wall region of a tank of a nuclear reactor installation such as is described, for example, in our commonly assigned U.S. Patent Application Publication US 2010/0192368 A1 and its corresponding German published patent application DE 10 2008 014 544 A1. Furthermore, the invention relates to a method for repairing such a damaged area. The wall surfaces (side walls and bottom surface) of water-flooded tanks of a nuclear reactor installation, for example the reactor pit or the fuel assembly storage tank, are provided with a liner made of steel plates welded to one another. The welds by means of which the steel plates are welded to one another or to a substructure are vulnerable to chlorine-induced stress crack corrosion as a result of mechanical stresses which inevitably occur during welding, with the result that cracks can occur over time. In order to prevent tank water from escaping into the concrete wall through such cracks, the cracks have to be sealed. In order to ensure sufficient shielding of the maintenance staff against radioactive radiation during such a repair, the water cannot be drained, in particular in the fuel assembly storage tank which is loaded with fuel assemblies, and therefore the repair has to take place under water. However, in such a fuel element storage tank in particular the side wall regions are not easily accessible since only a narrow gap is available between the fuel assembly storage rack, located in the fuel assembly storage tank, and the side walls. In principle, it is known, for example from the commonly assigned German patent application DE 100 26 649 A1, to close off such cracks by applying an adhesive or by adhesively bonding repair overlays onto them. For this purpose, the repair overlay was applied to the wall either manually by a diver or using a linkage system operated from the edge of the tank. However, in this way, it is not possible to repair damaged areas which are not easily accessible and are located at great depth underneath the water surface. In order to be able to repair even areas which are not easily accessible, our US 2010/0192368 A1 and DE 10 2008 014 544 A1 propose to arrange a guide system along a side wall, at a spacing distance therefrom, which guide system is secured to the side wall using suction cups and serves to guide a carriage which can be moved in a longitudinal direction of the guide system. A displaceably mounted receptacle for a repair overlay which can be applied with an adhesive surface to the wall region which contains the damaged area is arranged on the carriage. In this known device, the carriage is moved under its own weight into an end position in which the guide system is secured to the side wall with a particularly strong adhesive force using a plurality of suction cups in order to be able to absorb the opposing force which occurs when the repair overlay is pressed on. With the known device it is possible to reach side wall regions and edge regions at the bottom surface adjoining the side wall regions which are accessible only via a narrow gap due to fixtures located inside the tank, for example the fuel assembly storage rack of a fuel assembly storage tank. Owing to the relatively long curing time of the adhesive which is used, which amounts to at least twelve hours, using the known devices to repair extensive damaged areas, which may extend, for example, over the entire length of a vertically running weld seam, involves a large amount of expenditure in terms of time. It is accordingly an object of the invention to provide a device and a method for repairing a damaged area which overcome the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which provides for a device for repairing a damaged area in an underwater wall region of a container or tank, in particular in the wall region of a tank of a nuclear reactor installation, with which it is possible to repair a multiplicity of damaged areas which are not easily accessible or extensive, relatively large damaged areas with lower expenditure in terms of time. Furthermore, the invention is based on the object of specifying a method with which vertically extending weld seams can be quickly repaired over a relatively large longitudinal section. With the foregoing and other objects in view there is provided, in accordance with the invention, a device for repairing a damaged area of an underwater wall region of a container or tank, in particular a device for repairing a wall region of a tank in a nuclear reactor installation. The device comprises: a guide system to be mounted along a side wall, at a spacing distance from, and secured to, the side wall; at least one first carriage guided on the guide system, and movable in a longitudinal direction of the guide system; a receptacle displaceably mounted on the at least one first carriage, the receptacle being configured for holding a repair overlay to be applied with an adhesive surface to the wall region having the damaged area; and at least one suction mount disposed on the first carriage and configured for placement against the side wall and connected to a suction line. In other words, the first-mentioned object is achieved, according to the invention, with a device that contains a guide system which can be mounted along a side wall, at a distance therefrom, and can be secured thereto. The guide system has at least one first carriage which is guided thereon, can be moved in a longitudinal direction of the guide system and on which a receptacle, which is mounted so as to be displaceable perpendicularly with respect to this longitudinal direction, for a repair overlay, which can be applied with an adhesive surface to the wall region containing the damaged area, is arranged, wherein at least one suction mount, which can be placed against the side wall and is connected to a suction line, is arranged on the first carriage. By using a carriage which can be fitted onto the track of the guide system secured to the side wall and can be moved in the longitudinal direction of the guide rail, for the purpose of transporting the repair overlay to the damaged area it is possible to move a repair overlay provided with a viscous, free-flowing adhesive to the correct working position in a very short time since this has been previously defined by the mounting of the guide system. In other words: the mounting and positioning of the guide system can be carried out with a high level of precision and without time pressure since the repair overlay which is provided with the adhesive is not transported to the damaged area until the guide system has been installed. Since the first carriage itself is provided with at least one suction mount, the opposing force arising when the repair overlay is applied to the side wall can always be absorbed directly at the location of the carriage so that virtually no forces which are directed transversely with respect to its longitudinal direction and away from the side wall are applied to the guide system by pressing on the repair overlay. In this way, the first carriage can be stopped flexibly at different positions of the guide system and secured to the side wall without particular additional measures being required at these positions in order to securely fix the guide system to the wall, as is necessary in the device known from our above-mentioned US 2010/0192368 A1 and DE 10 2008 014 544 A1. As a result, a plurality of damaged areas at different longitudinal positions with respect to the secured guide system can be successively repaired without the guide system having to be repositioned. In one advantageous embodiment, the device comprises at least two first carriages and at least one second carriage which can be moved along the guide system and serves exclusively as a space keeper between first carriages which are adjacent to one another. Furthermore, if the distance between the repair overlays, which is brought about by the second carriage, corresponds approximately to the extent of a repair overlay in the longitudinal direction or is only slightly larger, in a second working step the non-covered gaps which are present between the individual repair overlays after the first step can be closed by using the same first and second carriages. If the first and second carriages are driveless and can be moved exclusively by the effect of gravity, the design of the device is significantly simplified. The assembly of the guide system in situ is facilitated if it is composed of sections which are detachably connected to one another. A particularly simple way of mounting the guide system on the side wall is possible if suction mounts which are connected to a suction line are arranged thereon. With the above and other objects in view is provided, in accordance with the invention, a method of repairing a damaged area of an underwater wall region of a container or tank, and particularly for repairing a wall region of a tank in a nuclear reactor installation. The method utilizes a device as summarized above and comprises the following steps: in a first working cycle, alternately fitting a multiplicity of first carriages equipped with repair overlays and second carriages onto the track of the guide system and bonding the repair overlays; and in a second working cycle, fitting the first carriages, again equipped with repair overlays, and the second carriages onto the track of the guide system in a reversed order relative to an order in the first working cycle. In other words, the second above-mentioned object is achieved in that, in a first working cycle, a multiplicity of first carriages that are equipped with repair overlays and second carriages, are alternately fitted onto the track of the guide system, and the repair overlays are bonded on. In a second working cycle, the first carriages, which have been equipped with repair overlays, and the second carriages are fitted onto the track of the guide system in a reversed order. In this way, a weld seam can be completely sealed over a long distance in only two working cycles. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a device and method for repairing a damaged area in an underwater wall region of a container or tank, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown a device according to the invention in a home position without first and second carriages in a water-flooded tank 2 of a nuclear power installation. The device comprises a supporting frame 4, held in a rack 3 positioned at the edge of the tank, for a guide system 6 which extends in a longitudinal direction 5, is constructed from two parallel guide rails 6a, b and is respectively composed of a multiplicity of sections 8 which are detachably connected to one another. A multiplicity of suction mounts 10 is arranged on the guide system 6 or on the guide rails 6a, b. It is possible to apply a partial vacuum to the suction mounts 10 via a suction line 10′, which is only schematically indicated. The suction mounts thus secure the guide system 6 vertically to a side wall 12 while being spaced apart therefrom at a spacing distance. In the example, the side wall 12 runs vertically. The wall surfaces (side walls 12 and bottom surface) of the tank 2 are provided with a liner 14 that is composed of steel panels welded to one another. In the illustrated example, the guide system 6 is arranged opposite a vertically running weld seam (not illustrated in FIG. 1) and extends as far as the bottom (likewise not illustrated) of the tank 2. The weld seam can have a multiplicity of damaged areas, with the result that it has to be sealed over its entire length in order to prevent tank water from penetrating the concrete wall 16. FIG. 2 shows that the guide system 6 is constructed from two guide rails 6a, b. The guide system 6 guides a multiplicity of first carriages 20 which are fitted onto the track thereof and each support a repair overlay 22. Between adjacent first carriages 20 there is in each case a second carriage 24 which serves as a space keeper and has approximately the same extent in the longitudinal direction 5 as the repair overlay 22. The first and second carriages 20, 24 are alternately fitted onto the track of the guide system 6 and moved downward with a cable winch on the supporting frame 4, exclusively under their own weight, i.e. as a result of the effect of gravity, wherein the carriages 20, 24 each form a stop for the next following carriage 24, 20. The device according to the invention is illustrated by way of example in more detail in FIGS. 2 to 5 on the basis of an embodiment which is provided for repairing a weld seam located in the corner region of a tank. This becomes clearer from the enlarged illustration according to FIG. 3. Every first carriage 20 has a receptacle 30 which is mounted on the first carriage 20 so as to be displaceable transversely with respect to the longitudinal direction 5 of the guide system 6, in an advancing direction 32 which is indicated by an arrow, and which holds the repair overlay 22. The receptacle 30 which is arranged on the first carriage 20 corresponds in its design substantially to the receptacle known from our above-mentioned US 2010/0192368 A1 and DE 10 2008 014 544 A1. As in the known device, the repair overlay 22 is secured in the receptacle 30 by generating a partial vacuum acting on a rear side facing away from the adhesive surface. This measure permits simple release or detachment of the repair overlay 22 when it is permanently secured to the wall after the adhesive has cured. In order to achieve sufficiently high pressing forces, the receptacle 30 is mounted so as to be pneumatically displaceable in the advancing direction 32 on the first carriage 20. However, in contrast with the embodiments explained in more detail in US 2010/0192368 A1 and DE 10 2008 014 544 A1, according to the invention the first carriage 20 itself is provided with a multiplicity of suction mounts 34 which can be placed against a side wall of the tank and which are connected to a suction line 34′. For use in the corner region, an angular profile, which is arranged on the correspondingly shaped receptacle 30, is provided as a repair overlay 22. The guide system 6 and the first carriages 20 are provided with suction mounts 10 and 34, respectively, which are oriented in pairs at 90° with respect to one another and with which the guide system 6 and respectively the first carriages 20 can be secured to two side walls which adjoin one another at a right angle. According to FIGS. 4 and 5, first and second carriages 20, 24 each have two runners 40a, b on which lateral rollers 42, with which the carriages 20, 24 roll along the guide rails 6a, b, are arranged. The narrow end faces 44a, b at the ends of the runners 40a, b each serve as a stop face for the carriage which is respectively adjacent in the stack, as can be seen clearly in FIG. 3. The distance between the repair overlays 30 of first carriages 20 which are adjacent to one another is approximately the same or at most slightly larger than the extent of the repair overlays 30 in the longitudinal direction 5. With the device according to the invention, a weld seam which extends in a longitudinal direction can be sealed virtually over its entire length in two working cycles. In a first working cycle, first and second carriages 20, 24 are alternately fitted onto the track, and the repair overlays 22 which are coated with adhesive are pneumatically pressed onto the wall. After the curing of the adhesive—approximately 12 hours—the receptacles 30 are moved back and the first and second carriages 20, 24 are successively removed from the guide system 6. The guide system 6 remains secured in an unchanged position to the side wall. In a second working cycle, the first carriages 20 are again provided with a repair overlay 22 which is coated with an adhesive, and the first and second carriages 20, 24 are alternately fitted, in a reverse order to that of the first working cycle, onto the track of the guide system 6 which is still secured to the side wall. The distance between the repair overlays 22 which is bonded onto the wall in the first working cycle is dimensioned such that a gap which, if appropriate, is located between adjacent repair overlays 22 in the second working cycle after the bonding of the repair overlays is covered by the adhesive which emerges laterally during the pressing-on process. In contrast to the exemplary embodiment illustrated in FIGS. 2 to 5, planar wall surfaces or cylindrical containers or pipes can also be repaired with correspondingly structurally adapted receptacles, guide systems and repair overlays. According to FIG. 6, a multiplicity of first carriages 202, 204 is fitted onto the track of the guide system 6, the receptacle 302, 304 of which is provided for repairing damaged areas on a planar wall surface with a planar repair overlay 220. The repair overlays 220 which are arranged on the first carriage 202 and which are arranged on receptacles 302 which extend in the longitudinal direction of the guide system 6 serve in this context for repairing weld seams which extend in a vertical direction. The carriage 204, which is illustrated in FIG. 6 and fitted onto the track on the guide system 6, supports a receptacle 304 which extends transversely with respect to the longitudinal direction 5 of the guide system 6. Accordingly, the planar repair overlay 220, which is also arranged thereon, serves to repair weld seams which run horizontally on a planar tank wall. In this exemplary embodiment, first carriages 202, 204 which are adjacent to one another are also spaced apart from one another in each case by a second carriage 24. In FIGS. 7 and 8, the first carriages 202 and 204, respectively, are each represented in an enlarged perspective illustration. The device according to the invention can also advantageously be used to repair an individual damaged area or damaged areas which are spaced apart from one another in the longitudinal direction of the guide system and which are located at different longitudinal positions of the guide system which is secured to the side wall with only a few suction mounts, since the reaction force which occurs when the repair overlay is pressed on is absorbed directly by the first carriage provided with suction mounts. |
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abstract | A system and method for reducing tritium migration. In one aspect, the invention is a method of reducing tritium mitigation from a spent nuclear fuel pool containing a body of tritiated water having an exposed surface, the method comprising hermetically sealing the exposed surface of the body of tritiated water with a cover movable between an open-state and a close-state. |
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summary | ||
047818859 | abstract | A nuclear reactor fuel assembly includes an elongated fuel channel with a square cross section and channel walls, the fuel channel having an imaginary lattice disposed therein with a box-shaped cross section having mesh openings and sides parallel to the channel walls, fuel rods containing nuclear fuel being mutually spaced apart in the mesh openings, and a prismatic water pipe spaced apart from the fuel channel by a given spacing and having a cross section spanning more than one of the mesh openings of the imaginary lattice, the given spacing being completely filled with the mesh openings of the imaginary lattice and the fuel rods disposed therein. |
053405064 | description | The following Examples are given to illustrate the invention, but are not to be taken as limiting the scope of the invention which is defined in the appended claims. EXAMPLE I Typical Preparation of the Sodalite Intermediate The sodalite intermediate was prepared by intimately mixing 2.8 g of NaOH, 32.8 g of Al.sub.2 O.sub.3, and 41.4 g of SiO.sub.2 (mole ratio of 2:1:2; weight ratio of about 1:1.27:6.60). About 100 g of this mixture is placed in a high-fired alumina crucible and heated to 500.degree. C. for 30 hours. The result is the reaction of the NaOH to form water, which is driven off, and compounds such as NaAlO.sub.2, Na.sub.2 SiO.sub.3, and Na.sub.2 Si.sub.2 O.sub.5, which are components of the sodalite intermediate along with Al.sub.2 O.sub.3 and SiO.sub.2. The products are kept dry and ground to a fine powder with particle sizes less than 500 .mu.m. The intermediate prepared in this manner is more reactive than a mixture of the pure materials. The reaction can be carried out in air, but the products are stored under a dry inert atmosphere, for example purified argon or helium. EXAMPLE II Immobilization of Zeolite-Salt Mixture A synthetic zeolite-salt waste material was prepared in a dry, inert atmosphere by mixing about 30 g of molten LiCl-KCl eutectic salt containing about 0.8 wt % SrCl.sub.2 2 wt % BaCl.sub.2, and 4.9 wt % CsCl with about 5 g of the sodium form of zeolite A. After gently mixing the salt and zeolite at 400.degree. C. for 8 hours, 21.8 g of the salt was separated by forcing it through a sintered steel filter having 50 .mu.m pores, and passing a stream of argon gas through the residue for about 1 hour. The filtered molten salt contained reduced amounts of strontium, cesium and barium, and negligible amounts of zeolite decomposition products. The salt-zeolite residue, which weighed 13.2 g, contained about 94% of the strontium, 85% of the barium, and 45% of the cesium that were in the original salt. The salt-zeolite residue was removed from the filter and ground to a powder with a particle size of less than 5005 m. EXAMPLE III Typical Immobilization of Zeolite-Salt Waste Material In a dry, inert atmosphere, 30 g of sodalite intermediate prepared as described in EXAMPLE I, and ground to a particle size of less than 500 .mu.m, was intimately mixed with 6.6 g of the zeolite-salt waste prepared as described in EXAMPLE II above. The resulting mixture contained sufficient intermediate to encapsule the 4.1 g of salt into sodalite if none of the zeolite converted to sodalite. This mixture was placed in a steel die, and heated to 325.degree. C. at 70 MPa for 6.5 hours. The resulting green pellet was sealed in a stainless steel container and heated for 750.degree. C. for 168 hours. The final pellet was hard and strong. EXAMPLE IV Typical Preparation of Sodalite from Waste Salts Crushed salt (LiCl-56 wt % KCl) was intimately mixed with sodalite intermediate prepared as described in EXAMPLE I in proportions that the salt content of the mixture is 10 wt %. The mixture is heated to 360.degree. C. in a steel die and pressed at 60 MPa. The preparation of a suitable green pellet is aided by the temperature being above the salt melting point. This pellet is then placed in a sealed stainless steel container and heated to 700.degree. C. for 100 hours to prepare a final pellet that is white, and very hard. EXAMPLE V The pellet prepared according to EXAMPLE IV was subjected to a leaching test similar to the standard procedure set forth in ANS 16.1. The pellet had a Leachability Index of about 13 for strontium and 12 for cesium indicating that the leach rates of these elements were about one-tenth their leach rates from mortars with the best formulations. |
054024544 | abstract | A process for obtaining a sample from an atmosphere in a closed gastight vessel, preferably from a reactor safety vessel of a nuclear power station, includes passing a sample through a venturi nozzle immediately upon entry of the sample into a sample-taking container in a vessel. The sample is mixed in the venturi nozzle with a transport fluid serving as a washing liquid. Gaseous constituents of the sample being soluble and/or condensable in the washing liquid are subsequently discharged together with the washing liquid from the sample-taking container and from the vessel by triggering a pressure reduction. A device for obtaining samples from an atmosphere in a closed gastight vessel, preferably from a reactor safety vessel of a nuclear power station, includes a sample-taking container having a bottom and a given volume. A washing liquid is disposed in the sample-taking container and has a volume being at most approximately equal to half of the given volume. A venturi nozzle dips into the washing liquid in the sample-taking container above the bottom. An inlet channel leads into the sample-taking container below the venturi nozzle. |
abstract | An ultraviolet area sterilizer or disinfector is incorporated into a building structure where concern exists regarding the presence of pathogenic bacteria on environmental surfaces. Ultraviolet C (UV-C) generators generate UV-C that is directed to architectural partitions of an enclosed area. The architectural partitions reflect UV-C to kill pathogens in the enclosed area. The device transmits a calculated dose of UV-C from a fixture mounted to an architectural partition in the enclosed area. Once an effective cumulative dose of UV-C has been reflected to radiation sensors, as measured by the sensors, the device shuts down. |
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description | This invention was developed under Contract DE-AC04-94AL85000 between Sandia Corporation and the U.S. Department of Energy. The U.S. Government has certain rights in this invention. The present work relates generally to neutron generators and, more particularly, to neutron generator designs that provide size scalability, ease of fabrication and multiple ion source functionalities. The utility of neutron generators in various endeavors is well known. Neutron generators are commonly used, for instance, in areas as diverse as oil well logging applications, and treatment/monitoring of medical conditions. Conventional high fluence, non-active, neutron generator technology is mostly based on vacuum accelerator or RF techniques. The most basic neutron generator uses high voltage to accelerate deuterium (D) ions. The accelerated ions impact on a metal target loaded with tritium (T) gas, causing a deuterium-tritium (DT) fusion reaction that produces neutrons. Such devices appeared in the literature in the early 1960's, and the design continues to evolve with variations on the accelerator type, power supply driver type, size, and output. A conventional neutron generator includes 1) a deuterium ion source, 2) an accelerating cavity, also termed an acceleration gap, or drift region, 3) an extraction plate disposed between the ion source region and the accelerating cavity, including an aperture for extracting the ions, and 4) the aforementioned metal target loaded with tritium. Most commercial deuterium ion sources are of the Penning type, which produces ions by heating a filament of wire (e.g., titanium) that has been hydrided with deuterium. As the temperature of the wire increases, the deuterium is released from the metal as a gas that is then ionized by a spark produced between a pair of electrodes. The deuterium ions are channeled through the aperture into the acceleration gap. At the end of the acceleration gap is the metal (e.g., titanium) target, which has been hydrided with tritium. The deuterium ions are accelerated across the gap by a high voltage applied between the extraction plate and the target. Conventional neutron generators typically use a cylindrically symmetric discharge geometry, and are thus commonly referred to as neutron tubes. The cylindrical geometry facilitates ion beam control and symmetrical radial beam expansion. This symmetric geometry, although simple and effective, is not easily scaled down, thereby disadvantageously limiting the possibilities of size reductions. It is desirable in view of the foregoing to provide a neutron generator that avoids disadvantages associated with prior art neutron generators. Exemplary embodiments of the present work provide a neutron generator having a flat, rectilinear geometry with surface mounted metallizations. This construction may be scaled down as desired, to as small as a micron size package, while maintaining a relatively high output. Some embodiments provide the same functional elements as in the conventional cylindrical geometry, but embodied in a flat arrangement of stacked dielectric layers that provide design flexibility at different neutron output levels. The ion source uses a pulse heating process to produce ions from flat strip metallizations deposited on dielectric substrates. Two electrodes face one another across a spark gap where an arc is produced to ionize deuterium gas. The electrodes also serve as filaments that release the deuterium gas when heated. As such, in some embodiments, the electrodes are made of titanium that has been hydrided or loaded with deuterium. A power source applies a high-voltage/high-current pulse to the electrodes, and the resulting joule heating releases the deuterium. The power source pulses with sufficient current and pulse width to produce the heating required to release the deuterium gas. The deuterium gas is released into a vacuum-sealed expansion cavity, and the pulsed voltage applied across the electrodes produces therebetween an arc to ionize the gas. The power source pulses have sufficient current and pulse width to sustain the arc so that the gas is ionized nearly simultaneously with its release. Accordingly, accumulation of background deuterium gas, which is known to limit system output in Penning-type ion sources, does not occur. The above-described dual use of electrodes as arcing elements for ionization and as filaments for releasing deuterium gas, as well as the above-described pulsing to produce the near simultaneous release and ionization of deuterium gas, are known from conventional neutron tube arrangements. However, the aforementioned use of flat strip metallization electrodes imposes power requirements commensurate with the flat strip construction. The required current and the required pulse length are readily calculated based on the thickness and length of the electrodes. For example, in various embodiments, the pulse width ranges in length from 10 nanoseconds to several microseconds, and the power source provides pulses in a range of 1 kV at 0.1 amps to 5 kV at 1 amp. The ionization operation produces in the expansion cavity an ion-rich plasma that expands toward an aperture in an extraction plate at one end of the expansion cavity. The aperture extracts the deuterium ions through operation of a voltage gradient provided by biasing the target to a higher voltage than the extraction plate. The aperture rejects electrons back into the plasma. The voltage gradient accelerates the extracted deuterium ions in an acceleration direction across an acceleration gap to the tritium-loaded target. When the deuterium ions impact the target, neutrons are produced by a conventional deuterium-tritium collision reaction. As shown in FIGS. 1A-1C, a neutron generator structure according to exemplary embodiments of the present work includes a plurality of dielectric substrate layers in a stacked arrangement. This stacked (i.e., layered or laminated) structure facilitates fabrication and size scaling. In some embodiments, all of the layers have a generally uniform rectangular size, shape and thickness, as shown in FIGS. 1A-1C. In various embodiments, the substrate layers are formed from a ceramic substrate material, a printed circuit board substrate material, or a semiconductor substrate material. In various embodiments, the layers range in thickness from approximately 0.5 mm to 3 mm and thicker. In some embodiments, the rectangular dimensions of the layers have approximately a 2:1 ratio, for example, 15×30 mm. The exploded view of FIGS. 1A-1C shows an example of five layers stacked successively upon one another. The stack structure of shown in the example of FIGS. 1A-1C includes a stack of three interior layers 1-3 disposed between two outer cover layers 4 that define opposite ends of the stack. In this example, the interior layers 1-3 have provided therein generally H-shaped openings that are substantially centrally located within the respective layers and substantially aligned within the laminate structure to produce a composite, H-shaped cavity having a height of three layers. As will be apparent from description which follows, the ion acceleration gap 22 (between extraction plate and target) is located in the “cross bar” portion of the H-shaped cavity 20. In various embodiments, the distance across the acceleration gap 22 (i.e., the width of the cross bar portion from left to right in FIG. 1) ranges from 3-20 mm. In some embodiments, the layers have rectangular dimensions of 15×30 mm, and the “legs” of the H-shaped openings have a length of approximately 12 mm. It will be evident to workers in the art that the H-shaped acceleration gap cavity 20 is provided as an expository example only. In various embodiments, openings of various different shapes provide the acceleration gap cavity 20 (see also FIG. 4). Layer 1, the middle layer of the stack structure, has provided therein a further opening 100 that is located adjacent the H-shaped opening of layer 1. In the example of FIGS. 1A-1C, the opening 100 is generally rectangular on three sides, and curved on the side adjacent the acceleration cavity. The opening 100 provides the plasma expansion cavity for the neutron generator. A relatively narrow notch 101 in the curved side of the opening 100 provides spatial communication between the acceleration cavity and the expansion cavity (see also FIG. 2). As described below, the notch 101 corresponds to the extraction plate aperture of the neutron generator. In some embodiments, the opening 100 has dimensions of approximately 3 mm in the left to right direction, and approximately 4 mm in the front to back direction. Some embodiments provide for a higher volume plasma expansion cavity (and correspondingly higher plasma densities and ion beam currents) by providing in interior layers 2 and 3 additional openings 100 that are aligned (in the stacking direction of the stacked structure) with the opening 100 of middle layer 1. Some embodiments provide one or more duplicates of layer 3. The volume of the acceleration gap cavity 20 is increased by providing such additional duplicate(s) of layer 3. Some embodiments provide one or more additional layers like layer 3, but also including an opening 100. The volumes of both the acceleration gap cavity 20 and the expansion cavity are increased by providing such additional layer(s). In various embodiments where the layers are approximately 15×30 mm rectangles, the acceleration gap 22 (i.e., the “cross-bar” portion of the composite H-shaped opening) has generally rectangular dimensions, as viewed in cross-section from left to right, that range from 3-9 mm in each direction. In various embodiments, the expansion cavity dimension in the stacking direction ranges from 1-9 mm. Operation of the neutron generator is controlled via five electrical terminals designated generally by 7 in FIGS. 1A-1C. Layer 1 has provided therein, at and end thereof adjacent the opening 100, four notches that respectively accommodate four of the electrical terminals 7. As described in detail below, the terminals accommodated by the notches in layer 1 are used to drive the ion source and bias the extraction plate. Layer 2 has provided therein, at an end thereof opposite the notches of layer 1, a notch that accommodates the fifth terminal 7. As described below, this terminal is used to bias the target. FIGS. 1A-1C further illustrate seven surface metallizations. In some embodiments, the metallizations are provided by conventional photolithographic techniques such as used, for example, in the fabrication of printed circuit boards. Three of the metallizations are longitudinal metallizations provided on longitudinal surfaces of the layers that face generally in the stacking direction of the stacked structure. Four of the metallizations are transverse metallizations provided on transverse surfaces of the layers that face generally transversely to the stacking direction. Transverse metallization 8 defines the target, for example, a tritium-loaded titanium metallization. The metallization 8 is provided on an elliptically curved transverse surface of middle layer 1 that faces across the ion acceleration gap 22 toward the opening 100. Longitudinal metallization 6 is an electrically conductive metallization provided on the longitudinal surface of layer 1 that faces layer 2. The metallization 6 electrically connects the target metallization 8 to the terminal 7 received in the notch of layer 2. Electrically conductive transverse metallizations 10 and 11 are respectively provided on elliptically curved transverse surfaces of layers 3 and 2 that are spatially aligned with one another in the stacking direction, and face across the acceleration gap cavity 20 toward the target metallization 8 on layer 1. Electrically conductive transverse metallization 9 is a two-part metallization. The component parts of metallization 9 are respectively provided on elliptically curved transverse surfaces of layer 1 that are adjacent notch 101. These elliptically curved transverse surfaces of layer 1 are spatially aligned in the stacking direction with the aforementioned elliptically curved transverse surfaces of layers 2 and 3, and face across the acceleration gap cavity 20 toward the target metallization 8. Each component part of metallization 9 includes at one end a generally hook-shaped portion that wraps around into the notch 101 such that each hook-shaped portion faces the other across the notch 101 (see also FIG. 2). The metallizations 9-11 contact one another by virtue of the layer stacking, and they define collectively the extraction plate of the neutron generator, as shown in FIG. 3. The wrap-around hook-shaped portions of the metallization 9 (shown at 301 in FIGS. 2 and 3) provide the aperture in the extraction plate. In various embodiments, the aperture width (i.e., the distance between the hook-shaped portions 301) ranges from 0.25-2 mm. Electrically conductive longitudinal metallizations 5 are respectively provided on longitudinal surfaces of layers 2 and 3 that face layer 1. Each of the two metallizations 5 is a three-part metallization, including parts 5B and 5C which electrically connect the aperture plate defined by metallizations 9-11 to respective ones of the terminals 7 received in the notches of layer 1. Each metallization 5 also forms a dual function ion source electrode/filament structure. As such, the metallizations 5 are (in some embodiments) constituted of titanium hydrided with deuterium. In particular, one part 5D of each metallization 5 extends into the expansion cavity defined by the opening 100 in the middle layer 1, and terminates in a generally hook-shaped portion 5E (see also FIG. 2). Each part 5B includes a structure that also extends into the expansion cavity, and terminates in a generally hook-shaped portion 5F. Each of the hook-shaped portions 5E and 5F of each metallization 5 provides the dual functions of deuterium release filament and ion source electrode. As such, the hook-shaped portions 5E and 5F of each metallization 5 operate as deuterium release filaments of the ion source, and also cooperate to form the electrode pair of the ion source. Each electrode of the pair terminates adjacent the other in the expansion cavity to define a spark gap 5G. The two metallizations 5 shown in FIGS. 1A-1C have generally the same structure, but are provided in opposite spatial orientations. More specifically, the orientation of the metallization 5 formed on layer 2 is flipped, i.e. rotated 180 degrees about a central longitudinal axis thereof, relative to the orientation of the metallization 5 formed on layer 3. Each spark gap 5G is located on a generally central longitudinal axis of the associated metallization 5, so the spark gaps 5G are generally aligned with one another in the stacking direction as shown in FIG. 2, despite the relative 180-degree rotation. It can be seen from FIGS. 1A-C and 2 that the aligned spark gaps 5G generally define a plane that extends left to right and in the stacking direction, and substantially bisects the extraction plate aperture 101. To reduce the incidence of metal particles from the arc material escaping from the expansion cavity into the acceleration cavity, some embodiments provide the spark gaps 5G laterally offset from the central longitudinal axis of the metallizations 5. As seen from FIGS. 1A-1C, the parts 5B and 5C respectively connect the extraction plate defined by metallizations 9-11 to opposite outer ones of the terminals 7 received in the notches of layer 1. These outer two terminals 7 are used to electrically control the extraction plate formed by metallizations 9-11. The parts 5D of the metallizations 5 on layers 2 and 3 are respectively connected to different ones of the inner two terminals 7 received in the notches of layer 1. Thus, one of the ion source electrode pairs is electrically controlled via one of the outer two terminals 7 and one of the inner two terminals 7, and the other of the ion source electrode pairs is controlled by the other of the outer two terminals and the other of the inner two terminals. A power source 15 drives the terminals 7 of FIGS. 1A-1C. Some embodiments ensure that the arcing across the spark gaps 5G may be produced by a relatively low voltage by, for example, setting the spark gaps 5G to a width of about 1 micron. As an example, some embodiments produce arcing with spark gap voltages ranging from 10-100 volts. As shown in FIG. 8, in some embodiments, the hook-shaped parts 5E and 5F are initially connected by a small bridging fuse 81 that burns away at the time of the first operation of the ion source. The fuse 81 is provided as a part of the metallization 5 that is significantly narrowed relative to the hook-shaped parts 5E and 5F. Although not shown to scale in FIG. 8, the fuse portion 81 is about 1/10 the width of the parts 5E and 5F in some embodiments, and 1/20 the width of parts 5E and 5F in other embodiments. The fuse portion 81 has various widths in various embodiments. The fuse portion 81 is narrow enough that the current density burns it away. Various embodiments having various numbers of ion sources are readily produced by supplementing an arrangement such as shown in FIGS. 1A-1C with additional layers to increase the volume of the expansion cavity (as explained above), and providing additional dual function electrode pairs on the additional layers. Although some embodiments provide a neutron generator having only a single ion source, two or more ion sources provide a number of advantages relative to single ion source embodiments. The following examples are illustrative. Some embodiments implement arc repetition by arcing the ion sources in alternating fashion, such that a sequence of arcs separated in time by a selected time interval occurs. This is shown for two ion sources in the simplified example of FIG. 5, wherein 51 graphically designates the arcing of an ion source. Some embodiments provide increased-power arcs by arcing two (or more) ion sources substantially simultaneously, as shown for two ion sources in the example of FIG. 6. Some embodiments provide a sequence of arcs analogous to the sequence of FIG. 5, but composed instead of increased-power arcs such as shown in FIG. 6. Some embodiments obtain an extended-length arc by arcing one ion source after, but in temporally overlapping relationship with, the arcing of another ion source, as shown for two ion sources in the example of FIG. 7. Some embodiments provide a sequence of arcs analogous to the sequence of FIG. 5, but composed instead of extended-length arcs such as shown in FIG. 7. Various embodiments that include multiple ion sources are capable of combining the techniques of FIGS. 6 and 7 to provide arcs having both increased power and extended length. Some embodiments provide a sequence of arcs analogous to that of FIG. 5, but composed instead of arcs having both increased power and extended length. The outer cover layers 4 of FIGS. 1A-1C are solid, without any openings. These cover layers 4 enclose the expansion cavity and the acceleration gap cavity 20, and permit those cavities to be vacuum-sealed. Various embodiments seal the laminate structure according to conventional techniques, for example, techniques such as used to seal stacks of printed circuit boards or stacks of integrated circuits. In some embodiments, the assembled neutron generator package is similar in size and shape to a computer chip socket. In some embodiments, the power source 15 is integrated into the laminate structure. As is evident from FIGS. 1A-1C, the ion acceleration gap 22 defined within the composite H-shaped cavity has a generally rectangular cross-section. As such, the ion beam should preferably have a cross section that is generally rectangular rather than cylindrical. A conventional cylindrical ion beam would spread over the lateral walls and cause high voltage breakdown. Shaping the ion beam in a rectangular cross section is similar, in the optical sense, to using a cylindrical lens to transform a circular cross section light beam into a rectangular, flat cross section beam. Exemplary embodiments of the present work provide an ion beam lens produced by the electric field distribution around the extraction plate aperture. This lens provides beam shaping to produce a generally flat, rectangular ion beam. Some embodiments implement the ion beam lens as follows. As indicated above, the metallizations 8-11 conform to the curvature of the corresponding elliptically shaped transverse surfaces of layers 1-3 on which they are deposited. Thus, the metallizations 8, 10 and 11 define ellipses, and the metallization 9 defines two truncated ellipses, each of which terminates in the hook-shaped portion 301 wrapping into the extraction plate aperture at 101. The parameters of the ellipses at 9-11 are related to the size and elliptical shape of the target metallization 8, and the ion acceleration gap 22 distance. The elliptically shaped extraction plate (9-11) is biased to ground or another fixed potential, and the elliptically shaped target 8 is biased to a much higher potential. For example, various embodiments bias the target 8 to various voltages ranging from 10 kV to 50 kV. This biasing of the elliptically shaped metallizations produces in the acceleration gap cavity 20 an electric field that tends to force the ion beam to be flat. The elliptically shaped extraction plate allows the ion beam to spread laterally as it proceeds toward the elliptically shaped target 8, thereby substantially covering the corresponding lateral (front to back in FIGS. 1A-1C) dimension of the target. The dimensions of the target 8, as well as the ion current and the target-to-extraction plate voltage, are dictated by the desired neutron output level. The required voltage dictates the acceleration gap 22 distance, and the required ion current dictates the width of the extraction plate aperture. After all dimensions are set, the extraction plate ellipse is designed such that the ion beam covers about 80% of both dimensions of the target metallization 8. Some embodiments incorporate a dimensional tolerance factor to reduce the likelihood that the ion beam will strike the cavity surfaces. In various embodiments, the dimensions of the target metallization 8 range from 1-10 mm in the stacking direction and 1-20 mm in the front to back direction of FIGS. 1A-1C, and the elliptical curvature of the target metallization 8 is tailored to those dimensions as a 2:1; 3:1, or 4:1 semi-major to semi-minor elliptical axis ratio. In various embodiments, the extraction plate (metallizations 9-11) has elliptical axis ratios similar to those of the target 8, but tailored to the acceleration gap 22 distance such that the beam does not spread beyond about 80% of the target metallization dimensions. In some embodiments, layers 1-3 are formed such that the conforming metallizations 9-11 have generally linear (straight) profiles as viewed in the stacking direction (rather than the elliptically-shaped profiles shown in FIGS. 1A-1C). In various embodiments, the stacking direction profiles of each metallization 9-11 has a generally linear central portion, but with curved portions at opposite ends of the linear portion. These curved portions of the profile present convex metal surfaces facing into the ion acceleration cavity. The size and shape of the convex surfaces may be tailored as desired to provide various corresponding focus options for the ion beam. For example, in some embodiments, the convex surfaces protrude forwardly (toward the target 8) from the central linear portion of the profile. Although the stacking direction profile of the two-part metallization 9 exhibits the aforementioned central linear portion in various embodiments, it retains the same curved central aperture structure shown at 301 in FIGS. 2 and 3. It will be evident that a complete set of neutron generator components, for example, the set shown in FIGS. 1A-1C, may be produced during a single fabrication run. As shown by the example of FIG. 4, some embodiments produce during a single fabrication run multiple complete sets of neutron generator components suitable for constructing multiple neutron generators. Component layers 1A-4A of FIG. 4 can be seen as generally corresponding to component layers 1-4, respectively, of FIGS. 1A-1C. Note that each of the multiple neutron generators shown in FIG. 4 is an example of aforementioned embodiments that have only a single ion source (i.e., only one of the spark gaps 5G of FIG. 2). In FIG. 4, because each of the neutron generators has only a single ion source, and because the portions 5B and 5C of adjacent metallizations 5 are commonly connected, the three illustrated neutron generators are controlled using only eight terminals 7. Thus, the control terminal to neutron generator ratio is 8:3 in FIG. 4, whereas the corresponding ratio in FIG. 1 is 5:1. In some embodiments, at least two of the multiple sets of components produced during a single fabrication run define respective neutron generators that differ from one another in at least one physical parameter (e.g., the acceleration gap distance). FIG. 9 illustrates operations that may be performed according to exemplary embodiments of the present work. At 91, an ion beam is produced, shaped to have a substantially rectilinear cross section. At 92, the ion beam is intersected with a metal target to produce neutrons. Although exemplary embodiments of the present work are described above in detail, this does not limit the scope of the work, which can be practiced in a variety of embodiments. |
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039714446 | description | DETAILED DESCRIPTION OF THE INVENTION In the drawing a reaction vessel, such as a nuclear reactor, is shown containing a charge or core made up of two groups of spherical elements 2, 3 arranged in uniform mixture. In each of the groups, the spherical elements have the same diameter, however, the spherical elements 2 in one group have a different diameter from the spherical elements 3 in the other group. Means, not shown, are arranged to circulate the uniformly mixed charge of spherical elements 2, 3 during the operation of the reaction vessel. The elements are introduced in through the top of the reaction vessel 1, passed downwardly through the vessel, and are removed through the discharge pipe located at the lower end of the vessel. Control rods 4, only one control rod is shown in the drawing, are positioned within the reaction vessel extending into the body of the uniformly mixed fuel elements for regulating the reaction. From the reaction vessel 1, each of the control rods extend upwardly into a cylinder 5. Each cylinder has a double-acting piston 6 at its upper end within the cylinder. By selectively supplying hydraulic or pneumatic fluid or pressurized gas (helium) into the cylinder, the control rods can be moved upwardly and downwardly, as required, within the reaction vessel 1. The diameter of the elements in each group is selected in such a way that the circulation of the elements 2, 3 within the reaction vessel does not cause any substantial segregation of the elements into separate groups. If, considering the reaction vessel, its parameters and the conditions to be experienced during operation, a standard size element could be selected as 60mm. With such a standard element forming the diameter for one of the groups, the diameter for the other group could be selected in the range of 5 to 35 % smaller or greater than 60 mm and preferably in the range of 5 to 20% smaller or greater. The arrangement of least two groups of fuel elements could be used in a reactor such as the type presently under construction in Germany known as the THTR (thorium high temperature reactor). The THTR is a 300 MWe nuclear power plant and uses helium as a coolant gas in the primary circuit for transferring the heat removed from the fission reaction to the steam generators in the secondary circuit. The reactor core is located within a prestressed concrete pressure vessel and has a diameter of 5.6m and a height of 6m. The core is made up of 675,000 fuel element spheres providing a pebble bed volume of 125m.sup.3. Each fuel element is 60mm in diameter, that is, about the size of a tennis ball, with an outer fuel-free graphite shell having a wall thickness of 5mm. Within the shell, the fuel is contained in the form of 35,000 small coated particles of 0.3 to 0.4mm diameter coated with pyro-graphite embedded in a graphite matrix. The heavy metal contents of each fuel element is 0.96g U235 (93 % enriched) and 962g Th 232. The core is enclosed within a graphite reflector with fuel element loading pipes passing through the upper end of the reflector for adding fuel element spheres to the core and with a centrally arranged fuel element discharge pipe located in the bottom of the reflector. To control and shut down reactor operation, 42 absorber or shutdown rods are arranged for insertion into the core in direct contact with the fuel elements with 36 control rods located laterally of the core in boreholes formed in the reflector. The incore rods are preferably designed for normal shut-down of the reactor as well as for scram in the case of fault conditions. The reflector rods are primarily used for temperature and partial load control. Continuous fueling of the core is practiced to ensure uniform and high burn-up of the fuel elements, the fuel elements move slowly downwardly through the core with the fuel elements in the central regions moving faster than those in the outer annular regions between the central region and the reflector. On the average a fuel element resides within the core a period of 36 months and usually is passed through the core five to seven times, preferably six times. As fuel elements are removed from the bottom of the core, they pass through the discharge pipe into a singulizer and are then led to a damaged sphere separator where fuel elements whose shape and dimensions have significantly changed are eliminated from the recirculation cycle. The other fuel elements, after having passed a buffer line, are led into a distinguishing and burn-up measurement device. According to sphere burn-up, and depending on the fueling program, a computer decides whether the element is discharged from the circulating cycle or recycled back into the core at its top. Continuous circulation means that during operation a certain number of elements per day at full power are added to the top of the core and a similar number are removed from the bottom. It is possible that a period of time may pass between each charging and discharging step, even up to several days in time, however, such circulation is still considered continuous. By comparison, it is known to leave fuel elements within a core for an extended period of time and then during a period of shutdown to replace at least a portion of the elements. However, such operation is not a continuous circulation of the elements moving downwardly through the core during reactor operation. The helium gas coolant is forced downwardly through the reactor core, removing heat from the fuel elements and exiting through boreholes in the bottom reflector. If, instead of a single size fuel element, the present invention is employed and two groups of differently sized elements are utilized, it would be preferable to take the 60mm diameter of the THTR fuel elements as a standard. To assure a random rather than ordered orientation of the fuel elements within the core, as might develop when a single size element is used, at least one additional group of fuel elements is selected for use along with a group of the 60mm elements. The additional group would be in the range of 5 to 35 %, and preferably in the range of 5 to 20 %, greater or less than the standard 60mm diameter. In other words, the diameter of one group would be 60mm while the diameter of the other group would be selected from the range of 48 to 57mm or 63 to 72mm, depending on whether, in accordance with the various conditions involved, it is decided to employ a fuel element size smaller or larger than the standard size. To determine the difference in pressure acting on control rod fuel elements when one size fuel element is employed in the core as compared with two groups of fuel elements each of a different diametral size, testing has indicated that a very significant reduction in fuel element stress is obtained where two different sized diameter spherical elements are used and are uniformly mixed together. It had been thought that the use of two different size elements would result in segregation due to continuous recycling, however, by employing the two different diametral sized groups, it is possible to maintain a random and non-segregated arrangement which assures continuous recirculating operation with adequate control without adverse stress effects on the fuel elements and the control rods. While a specific embodiment of the invention has been shown and described in detail to illustrate the application of the inventive principles, it will be understood that the invention may be embodied otherwise without departing from such principles. |
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