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abstract | Disclosed is a control rod drive mechanism. More specifically, the control rod drive mechanism includes a guide member 100 disposed in a nuclear reactor to receiving a drive shaft 2; a latch assembly 200 disposed in the guide member 100 to enable the drive shaft 2 to be withdrawn and inserted; a supporting member 300 connected to the guide member 100 to cover the drive shaft 2 and to support the latch assembly 200; and a plurality of coil housings 400 spaced apart and connected to the guide member 100 to cover the latch assembly 200, and each having a coil 410 built therein. |
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058959190 | description | A gun lens 1 for generating a particle beam is shown in FIG. 1. It essentially comprises a cathode 2, an extraction electrode 3, an anode 4 and a condenser lens 5. The cathode 2 is constructed as a photocathode or a field emission cathode as required. It is supplied with a potential U.sub.K. Furthermore, the extraction electrode 3 and the anode 4 are at a potential U.sub.Ex or U.sub.An respectively. The potential distribution between the cathode, extraction electrode and anode is chosen so that a deceleration field is generated between the extraction electrode 3 and the anode 4. In this deceleration field the particles emitted on the cathode 2 are preferably decelerated to a final energy of less than 3 keV. The condenser lens 5 generates a magnetic field which is superimposed on both the cathode 2, the extraction electrode 3 and the anode 4. An axial magnetic field distribution is indicated by way of example in FIG. 1 and is designated by the reference numeral 6. The condenser lens 5 conventionally consists of an iron circuit 5a and a coil 5b. In the illustrated embodiment the condenser lens 5 is constructed as a single-pole lens. Especially when the cathode 2 is constructed as a thermal field emission cathode it is particularly advantageous to superimpose this cathode with a suppressor electrode 7. This suppressor electrode 7 suppresses the particles which are not emerging directly from the cathode tip. Particles which originate from areas remote from the tip expand the virtual source size and accordingly reduce the brightness. In FIG. 2 a second embodiment of a gun lens 1' is illustrated, which differs from the embodiment according to FIG. 1 only by an element made from magnetic material 8 for influencing the magnetic field generated by the condenser lens in the region of the cathode. This element 8 made from magnetic material is advantageously disposed between the cathode 2 and the extraction electrode 3. In the illustrated embodiment according to FIG. 2 the element 8 made from magnetic material is coupled to the iron circuit 5a of the condenser lens 5. With this magnetic element 8 the magnetic field of the condenser lens 5 can be precisely influenced in the region of the cathode tip. Within the scope of the invention the magnetic element 8 and the extraction electrode 3 can be constructed as one component which takes on the function of the extraction electrode and also of the magnetic element. However, it would also be conceivable for the element 8 to contain the extraction electrode 3 as a diaphragm insert, in which case both electrodes could be at the same or different potential. A particle beam device with a gun lens 1" and a beam blanking system 9 is illustrated in FIG. 3. The gun lens 1" differs from the gun lens according to FIG. 2 again only by an element 8' made from magnetic material which influences the magnetic field of the condenser lens 5. The element 8' in this case is not coupled to the iron circuit 5a of the condenser lens 5 but, rather, forms an air gap with the iron circuit 5a. As the individual electrodes in the gun lens are charged so that the emitted particles are already decelerated to their final energy in the gun and moreover the anode is at ground potential, all subsequent components of the particle beam device can also be grounded. Since the particle beam is usually guided in a vacuum, this makes it possible to dispense with the costly decoupling of the high voltage from subsequent components. Apart from other diaphragms and lenses the particle beam device also has in particular a beam blanking system 9 with which the particle beam striking a specimen 10 can be blanked if required. For a high spatial resolution in the region of the specimen it is particularly advantageous if the gun lens produces a beam crossover in the region of the beam blanking system 9. Thus the gun lens also acts as a focusing element which images the virtual beam crossover of the gun into the optical system. The focal length with a given final energy of the particle beam is determined by the potential of the extraction electrode 3 and the excitation of the condenser lens 5. In order to achieve constant lens properties, i.e. a constant focal length, if the potential U.sub.Ex is altered the excitation of the condenser lens 5 must be adjusted correspondingly. The control of the potentials as well as the excitation of the condenser lens is effected by way of a control unit 11 which operates according to a predetermined program. At a final energy of the particles of in the region of 200 eV to 5 keV the voltage of the extraction electrode 3 is in the range between 3 kV and 7 kV, the voltage depending essentially upon the geometry of the source, the distance of the extraction electrode 3 from the emitter and the emission current. A typical voltage of the suppressor electrode 7 is between -100 Volt and -1 kV. The gun lens 1, 1', 1" according to the invention is distinguished by high brightness, low aberration coefficients and a high spatial resolution. |
052992438 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS As illustrated in FIG. 1, the glove holder unit according to the invention equips a confinement enclosure 1 having walls 2 separating the enclosure exterior 3 from the interior 4. In conventional manner, said enclosure 1 also has a filter 5, a pump 6 ensuring a vacuum within said enclosure and a transfer box 7 making it possible to insert or remove tools or products with respect to the interior of the enclosure 1. One of the enclosure walls 2 also has a transparent window 8 enabling the user to look into the interior of the enclosure 1. As illustrated in FIG. 4, the wall 2 has at least one access opening 9 (generally there are two for the user's two arms), said opening preferably being circular. A glove disk or ring 10 is fixed at said wall opening 9. A sleeve 11 for protecting the arm and the forearm, as well as a glove protecting the hand are fixed in detachable manner to said glove disk 10. The protective sleeve 11 comprises a first part 12 equipped with bellows and a second part 13 made from a flexible material. In addition, a working glove 15 is fixed by detachable assembly means 17 to the second part 13 of the protective sleeve. More specifically, the first part 12 equipped with bellows makes it possible to protect the arm, while the flexible part 13 protects the forearm and the elbow. This facilitates the movements of the user. The respective lengths of said two parts 12, 13 correspond to the average dimensions of the arm of a user. Advantageously, the rigid, bellows-equipped part 12 is made from semi rigid synthetic rubber and the flexible part 13 is made from vinyl. However, it is also possible to choose other materials provided that they comply with the requisite rigidity and flexibility criteria and have a resistance to oxidation and/or radiolysis phenomena and/or strong aciss. The assembly means 17 of the glove 15 and the sleeve 11 comprises a glove holder bracelet fixed to the end of the flexible part 13 of the sleeve and able to receive the glove 15. The glove holder bracelet 17 is constituted by a ring or a cylindrical tube portion having on its outer surface two substantially, annular grooves 19 and 21, the groove 19 being located substantially parallel in the center of said bracelet. These grooves 19, 21 make it possible to receive an O-ring 22 integral with the end of the glove 15. Moreover, said glove holder bracelet 17 has at its end opposite to that with the groove 21, a truncated coneshaped part 23 to which is fixed the end of the sleeve part 13. Said fixing can take place by bonding or heat sealing. The fixing is such that the glove holder bracelet 17 is located within the flexible part 13 of the sleeve. The glove disk 10 also has a cylindrical tubular shape. It has a peripheral, annular rib 25 defined by two annular ledges or rims 27, 29. The rim 27 is placed in the interior 4 of the confinement enclosure 1, while the rim 29 is placed on the exterior 3 of the enclosure. The rib 25 receives the edges of the wall 2 forming the opening 9. The glove disk 10 also has a lateral flange 31 extending towards the outside of the enclosure perpendicular to the wall 2. This annular, lateral flange 31 also has on its outer periphery two annular grooves 33, 35 respectively. The groove 33 is located between the groove 35 and the rib 25. The groove 35 receives the O-ring 37 forming the end of the bellows-equipped sleeve 12 and thus ensures the detachable fixing of said sleeve. Means 39 for sealing the first sleeve part 12 is provided and positioned at the junction between the two protective sleeve parts 12 and 13. More specifically, said sealing means 39 comprises a ring which is threaded on its outer surface and fixed by bonding or heat sealing respectively to the free end of the bellows-equipped sleeve 12 and the free end of the flexible part 13. The thread is designated 40. This ring 39 is able to cooperate with a sealing or closing plug 41 shown in FIG. 7 and which will now be described in greater detail. This plug 41 is constituted by a disk or base 42 made from a plastic material and having a tapped, peripheral, annular rim 43 so that it can be screwed onto the thread 40 of the ring 39. The tap of the plug 41 constitutes an example of fixing means to the ring 39. However, it is obvious that other fixing means such as bayonet fixing means, a catch or force assembly can be used without passing outside the scope of the invention. In addition, said plug 41 is provided substantially in its center with an aerating grid or grating 45. It also has a gripping handle 47 fixed to the plug base 42 on either side of the grid 45. When the plug 41 is in place on the ring 39, said grid 45 ensures the vacuum within the bellows 12. The glove holder unit according to the invention also has means for locking the assembly means 17 of the glove 15 and the sleeve 11. These locking means is more particularly illustrated in FIGS. 5 and 6 and carries the general reference numeral 49. The locking means 49 preferably has two substantially identical, symmetrical elements, but there can also be more than two elements. The upper element (relative to FIG. 6 and the normal use direction), carries the reference numeral 51 and the lower element the reference numeral 53. The elements 51, 53 are placed on the lateral flange 31 of the glove disk 10 so as to cover the latter. More specifically, each element 51, 53 is shaped like a semicircularly curved channel with a substantially C-shaped section, as can best be gathered from FIG. 5. This channel is positioned "astride" on at least one part of the lateral flange 31 of the glove disk. Therefore each element 51, 53 comprises a lateral, peripheral edge 55 terminated by an element forming a gasket 57 for engaging in the groove 33 of the glove disk 10. This edge or rim 55 is extended towards its center by a front face 59 and by a slightly oblique portion 63. The latter is internally extended by a circular arc-shaped shoulder 65 for receiving the truncated cone-shaped part 23 of the glove holder bracelet 17. Moreover, said two elements 51, 53 are provided internally, level with their central opening 67, with four return-preventing strips 69 made from a thick, flexible plastic material (cf. FIG. 6). These strips 69 prevent the glove 15 from being sucked into the interior of the bellows-equipped protective part 12. The locking means 49 can occupy two positions, a first locked position in which the two elements 51, 53 are locked by means 71 and a second unlocked position, in which these two elements are spaced apart from one another and permit the passage of the glove disk 17 and the glove 15. The locking means 71 constituted by two spring hooks provided on either side of the two elements 51 and 53. In the locked position, the two elements 51 and 53 are engaged against one another (left-hand part of FIG. 6) and in the unlocked position, the lower element 53 is slightly spaced from the upper element 51 (right-hand part of FIG. 6). The operation of the device will now be described in greater detail relative to FIGS. 2A to 3C. FIG. 2A illustrates the glove holder unit in the inoperative position, i.e. when the user's hand and arm are not in the interior. It should be noted that the locking means 49 is only partly shown. When the user wishes to change the glove 15, whose O-ring 22 is positioned in the annular groove 21, he introduces his hand and arm into the glove holder unit in the manner shown in FIG. 2B and seizes the sealing plug 41 by means of its gripping handle 47. He withdraws his hand and arm into the protective sleeve 11 until the plug 41 is level with the ring 39 and then screws the plug onto said ring by a quarter turn rotation. He is then in the position illustrated in FIG. 2C. Then, he continues to withdraw his hand and the glove 15 until the latter is outside the enclosure and turns over the glove 15 in the manner illustrated in FIG. 3A. In order to do this, it is obviously necessary for the locking means 49 to be unlocked (bottom position of the element 53 illustrated in FIG. 2C). Then, using his other hand, he locks the locking means 49 and the ring 17 consequently rests on the shoulder 65 of the locking means 49. He is then able to slip over the worn glove 15, which is inside out, a new glove 15', which is also inside out and whose O-ring 22' is located in the annular groove 19 FIG. 3A. He then inserts his hand in the new glove 15' and seizes the worn glove 15, removes the O-ring 22 of the glove 15 from the annular groove 21 and drops the worn glove 15 into the glove holder unit. The O-ring 22' of the new glove 15' is then moved from the annular groove 19 into the annular groove 21, while maintaining maximum contact with the glove holder bracelet 17. In order to bring about maximum sealing and security, at least one turn of the not shown adhesive tape is placed at the annular groove 21 in which is placed the O-ring 22'. Then, as illustrated in FIG. 3B, he again inserts his hand in the glove 15', so as to put it into place within the bellows 11. He can then unlock the plug 41 by rotating the latter and place it within the enclosure. He is then in the situation shown in FIG. 3C, where his hand is protected by the glove 15'. |
H00004073 | claims | 1. Apparatus for production of an electrical voltage, the apparatus comprising: a hollow, closed electrically conducting sphere of inner radius substantially greater than one meter, the sphere having a spherical region at the sphere center removed to allow positioning of a laser fusion target at the sphere center, and the sphere wall material having an associated electron work function of substantially 1.6 eV; one or more thin, electrically conducting anode plates positioned in the sphere interior adjacent to and spaced apart from and substantially parallel to the adjacent sphere wall, each such plate having an associated electron work function that is >>1.6 eV; an impedance or other electrical load connecting each anode plate and the adjacent sphere wall; an inert gas, capable of forming excimers and maintained at a pressure of substantially 100 atmospheres, contained in the interior of the hollow sphere; a fusion target positioned substantially at the center of the sphere; and a laser beam, focused on the last fusion target. a hollow, closed container having at least one substantially planar wall containing Li.sup.6 and containing electrically conducting material with an associated electron work function .PHI..sub.c.apprxeq. 1.6-6 eV, with the planar wall serving as a cathode; a screen anode, positioned in the container interior at a position spaced apart from and substantially parallel to the planar wall, the anode material containing electrically conducting material; an impedance or other electrical load connecting the cathode and the anode; an inert gas, capable of forming excimers and maintained at a pressure of substantially 1-30 atmospheres, filling the container interior; and a source of energetic neutrons of energy substantially 10 MeV or higher, positioned outside the container and adjacent to the planar wall. further including a second impedance or other electrical load connecting said screen anode and the second cathode. providing a hollow, closed container having at least one substantially planar wall that contains Li.sup.6 and contains electrically conducting material with an associated electron work function of .PHI..sub.c .apprxeq.1.6-6 eV, with the planar wall being adapted to serve as a cathode; providing a screen anode in the container interior at a position spaced apart from and substantially parallel to the planar wall, with the screen anode material being a metallic or other electrical conductor; providing an impedance or other electrical load connecting the planar wall and the screen anode; providing an inert gas of particles X of He, Ne or Ar, maintained at a pressure of 1-30 atmospheres, in the container interior; directing a beam of energetic neutrons of energy substantially 10 MeV or higher at the planar wall of the container; allowing the energetic neutrons to collide with Li.sup.6 particles in the substantially planar wall to produce high energy helium ions by reactions such as n+Li.sup.6 .fwdarw.He.sup.++ +He.sup.++ +(H.sup.3).sup.+ +3e.sup.- ; allowing the high energy helium ions produced by the n+Li.sup.6 reactions to react with the inert gas particles X to produce ions and excimers and short wavelength radiation of wavelength .lambda..sub.d =c/.nu..sub.d by excimer reactions such as EQU X.sup.+ +2X.fwdarw.X.sub.2 +X, EQU X.sub.2 +e.sup.-.fwdarw.X*+X, EQU X*+2X.fwdarw.X.sub.2 *+X, ##STR2## allowing a portion of the dissociation radiation of wavelength .nu..sub.d to strike the adjacent planar wall and eject photoelectrons of energy substantially 10-20 eV; and allowing the photoelectrons ejected by the planar wall to move to the screen anode and create an electric potential difference between the planar wall and the screen anode. providing a hollow, closed sphere of radius substantially greater than one meter, the sphere having a region at the sphere center removed to allow positioning of a laser function target at the sphere center, the sphere material being electrically conducting and having an associated electron work function of substantially 1.6 eV and having wall thickness sufficient to withstand an interior gas pressure of substantially 100 atmospheres, with at least a portion of the sphere wall being adapted to serve as a cathode; providing one or more electrically conducting anode plates that are optionally thin to electromagnetic radiation at the wavelength .lambda..sub.d, positioned in the sphere interior adjacent to and spaced apart from and substantially parallel to the adjacent sphere wall, each anode plate having an associated electron work function .PHI..sub.a <<1.6 eV; providing an impedance or other electrical load connecting the anode plate and the adjacent sphere wall; providing an inert gas of particles X of He or Ne in the sphere interior at a pressure of substantially 100 atmospheres; providing a laser fusion target substantially at the sphere center; irradiating the fusion target to produce a plurality of high energy neutrons of energy substantially 10 MeV or higher; allowing the high energy neutrons to move through and collide with the inert gas particles X to produce ions and excimers and electromagnetic radiation of wavelength substantially .lambda.=.lambda..sub.d =c/.nu..sub.d, where c is the velocity of light, by reactions such as EQU n(fast)+X.fwdarw.n(fast)+X.sup.+ (fast)+e.sup.-, EQU n(fast)+X.fwdarw.n(fast)+X.sup.++(fast)+ 2e.sup.-, EQU X.sup.++(fast)+X.fwdarw.X.sup.++(fast)+X.sup.+ +e.sup.-, EQU X.sup.+ (fast)+X.fwdarw.X.sup.+(fast)+X.sup.+ +e.sup.-, EQU X.sup.++ (fast)+X.fwdarw.X.sup.++(fast)+X.sup.*, EQU X.sup.*+ 2X.fwdarw.X.sub.2.sup.* +X.fwdarw.3X+h.nu..sub.d, EQU X.sup.+ +2X.fwdarw.X.sub.2.sup.+ +X, EQU X.sub.2.sup.+ +e.sup.- .fwdarw.X.sub.2.sup.* .fwdarw.2X+h.nu..sub.d ; allowing the electromagnetic radiation thus produced to pass through one or more of the anode plates and strike the sphere wall to produce a plurality of photoelectrons; and allowing the photoelectrons to move from the sphere wall to the anode plate, thus producing an electrical potential difference between cathode and anode plates. 2. Apparatus according to claim 1, wherein said inert gas is drawn from the class consisting of He and Ne. 3. Apparatus for production of an electrical voltage and of electromagnetic radiation of wavelength substantially 640 .ANG. and higher, the apparatus comprising: 4. Apparatus according to claim 3, wherein said hollow closed container further includes a second substantially planar wall containing Li.sup.6 and a metal or other electrical conductor with an associated work function .PHI..sub.c .apprxeq.=1.6-6 eV, with the second planar wall also serving as a cathode, with the second planar wall being spaced apart from and substantially parallel to said first planar wall and said screen anode so that said screen anode is positioned between said first planar wall and the second planar wall; and 5. Apparatus according to claim 3, wherein said inert gas is drawn from a class consisting of He, Ne and Ar. 6. A method for generation of an electrical voltage and of electromagnetic radiation of wavelength substantially 640 .ANG. or higher, the method comprising the steps of: 7. A method for generation of an electrical voltage and for production of electromagnetic radiation of wavelength .lambda..sub.d that is substantially 640 .ANG. or higher, the method comprising the steps of: |
050948024 | abstract | A nuclear fuel assembly with a debris filter. The lower end fitting of the fuel assembly has a stamped plate attached thereto that serves as a debris filter immediately upstream of the fuel rods. The stamped plate is provided with a plurality of flow holes in a size and pattern that provides filtration of debris damaging to the fuel rods while maintaining adequate coolant flow through the fuel assembly. |
claims | 1. A canister that is capable of storing and transporting spent nuclear fuel, the canister comprising:a canister shell that includes an open end;a closure lid that is inserted within the open end of the canister shell, the closure lid having a sidewall, an upper portion having a first outer lateral circumference and a lower portion having a second outer lateral circumference, the first outer lateral circumference less than the second outer later circumference;a first weld layer disposed in a weld area, the first weld layer having welds between the closure lid and the canister shell at the top of the lower portion of the closure lid, the first weld layer being configured to close the closure lid onto the canister as a first seal to the canister;a second weld layer disposed above the first weld layer, the second weld layer having welds between the closure lid and the canister shell and disposed on the first weld layer, the second weld layer being configured to close the closure lid onto the canister as a second seal to the canister; anda seal ring that is disposed above the second weld layer and between the canister shell and the upper outer circumference of the sidewall of the canister lid, a lower portion of the seal ring being placed substantially adjacent to the second weld layer, a top surface of the seal ring being substantially aligned and welded to the top surface of the closure lid and a top end of the canister shell, the seal ring being configured to close the closure lid onto the canister as a second seal to the canister. 2. The canister as defined in claim 1, wherein the at first weld layer and the second weld layer have a composite weld layer depth to assure closure of the canister. 3. The canister as defined in claim 1, further comprising a basket assembly that is disposed in the canister shell. 4. The canister as defined in claim 2, wherein the first weld layer and second weld layer are cooled below a material-specific adhesion temperature after the each of the weld layers are applied to assure each of the weld layers has achieved adhesion between the canister shell and closure lid. 5. The canister as defined in claim 4, wherein the first weld layer and second weld layer are inspected after a prescribed number of weld passes is cooled using approved methods and techniques to determine the quality of the weld layers and their acceptability, the approved methods and techniques including a dye-penetrant examination. 6. The canister as defined in claim 1, further comprising weld surfaces that weld the seal ring to the closure lid and to the canister shell. 7. The canister as defined in claim 6, wherein weld surfaces are applied as single pass welds or multi-pass welds. 8. The canister as defined in claim 6, wherein the weld surfaces are cooled below a material-specific adhesion temperature after the weld surfaces are applied to assure the weld surfaces have achieved adhesion among the seal ring, canister shell and closure lid. 9. The canister as defined in claim 8, wherein the weld surfaces are inspected after the weld surfaces are cooled using approved methods and techniques to determine the quality of the weld surfaces and their acceptability, the approved methods and techniques including a dye-penetrant examination. |
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abstract | Apparatus for generating ultraviolet light and methods of operating an ultraviolet light source. The apparatus may include a microwave chamber (16) enclosing an interior space, a light source (10) with a lamp head (28) coupled to the microwave chamber (16), an ultraviolet (UV) transmissive member (88) positioned above the lamp face (32) and below the interior space to define a plenum (116) therebetween, and an exhaust system (100) coupled in fluid communication with the plenum. The lamp head (28) has a lamp face (32) through which ultraviolet light (34) and cooling air (30) are emitted. The UV transmissive member (88) is configured to transmit the ultraviolet light (34) into the interior space and to divert the cooling air (30) from the interior space. The exhaust system (100) configured to exhaust the cooling air (30) from the plenum (116). |
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041586395 | description | Referring now to the drawing, there is shown an autoclave 1 having cylindrical sidewalls 2 and endpieces 3 and 4. (It should be understood) that the apparatus described herein is only exemplary of apparatus suitable for practice of the claimed methods. Numerous autoclave designs exist which, with some modifications can be made suitable for the practice of this invention.) The endpieces are held to flanges attached to the cylindrical sidewall by suitable fasteners. Seals 5 in the nature of O-rings are provided to insure that the vessel is pressure tight. Within the autoclave is an insulating hood 6. Preferably, the hood has a removable top cover. Electrical heating elements 7 are positioned along the inside of the hood. In the center of the vessel is a pedestal 8. The vessel endpiece 4 is provided with suitable lead-throughs (not illustrated) for the power lines to the heating elements and for thermocouples for measuring the temperature for control purposes. A pressurizing system for the vessel comprises a valve 9, a pump 10 and reservoir 11 for introducing high pressure gases into the vessel. What has been described to this point is an autoclave of the type used for hot isostatic pressing. The gas in the reservoir is usually argon to protect the electrical heating elements from oxidation. A pressure sensor 12 in communication with the interior of the autoclave provides an electrical output signal indicative of the pressure in the vessel. Adjacent the high pressure vessel is a safety container 20 which has a removable cover 21. The safety container has a gas tank 22 for holding the supply gas or gas to be stored. It also contains a bidirectional pump 26 and valve 23. A conduit 24 passes between valve 23 and into the pressure vessel. A pressure sensor 25 is in communication with the conduit 24 which sensor provides an electrical output indicative of the pressure in the conduit 24. According to this process. a canister is filled with a capturing solid 37 (shown in the breakaway portion of the canister 30) and is placed upon the pedestal 8 while the autoclave cover 3 and the hood cover are removed. The canister is provided with a sealable cover 31 to which is secured a gas valve 32 which when opened enables the interior of the canister to communicate with a fitting 33. A nipple 34 enables the fitting 33 to be connected to the conduit 24. Where the particular process is the encapsulation of radioactive krypton, the canister 30 is filled with a zeolite. The canister is connected as shown in the drawing with the valve 32 in the open position. The autoclave is then closed and bolted. At this time, the canister 30 and the capturing solids 37 are heated within the hood 6 by the heating elements 7. When the temperatures is appropriate to absorption (say, over 500.degree. C.), valves 9 and 23 are open and pumps 10 and 26 are activated so that the pressure of the vessel and the pressure within the canister are simultaneously raised. These pressures are monitored by the sensors 12 and 25 and any pressure differential is recognized by controller 40 which controls the pumps and valves to minimize the pressure difference across the canister wall. The pressures contemplated are in excess of 15,000 psi. At some time the pump 10 may stop and valve 9 close while the pump 22 continues to introduce krypton into the canister 30 as it is absorbed therein. When the krypton can no longer be introduced in the canister without raising the pressure thereof the capturing solid is considered loaded. At that time, the vessel is allowed to cool and the vessel and canister are evacuated maintaining the pressure differential constant until the atmospheric pressures are achieved inside and out of the canister. Then the autoclave cover is open and the valve 32 closed. Nipple 34 and conduit 24 can then be completely evacuated by pump 26 and valve 23 may be closed. The nipple 34 is then disconnected. The canister is ready for storage. In this way, neither of the atmosphere or the inside of the autoclave is ever exposed to radioactive krypton. The advantages according to this invention result from the canister 30 having its own pressure connection protruding through the main vessel closure and thereby during pressurization only the canister internals are exposed to radioactive material. The fact that the canister 30 may be made from thin, that is, relatively nonpressure resistant materials is a considerable advantage in that it permits quick heat up and cool down. Having thus described my invention with the detail and particularity required by the Patent Laws, what is desired protection by Letters Patent is set forth in the following claims. |
claims | 1. A basket for transport and/or storage packaging of radioactive materials, the basket comprising:at least one internal partition comprising at least one wall having two opposite lateral surfaces, andat least one peripheral partition,the internal partition delimits at least partially on either side of the same two cells intended to house the radioactive materials, the peripheral partition participating with the internal partition in delimiting the cells,the peripheral partition comprising at least one housing accommodating one end of the at least one wall, the housing comprising two opposite lateral housing surfaces and a bottom bringing together the two lateral housing surfaces,wherein the basket comprises a fastener configured to press at least one of the lateral wall surfaces against at least one of the lateral housing surfaces and to tighten the lateral wall surface against the lateral housing surface. 2. The basket according to claim 1, wherein the internal partition comprises two parallel walls separated by a spacing, each of the walls having a lateral external surface and a lateral internal surface,wherein at least one of the lateral external surfaces is configured to be pressed against one of the lateral housing surfaces by the fastener, orwherein at least one of the lateral internal surfaces is configured to be pressed against one of the lateral housing surfaces by the fastener. 3. The basket according to claim 1, wherein the fastener generates a pinching strain of the wall end of the internal partition between:said lateral housing surface against which the wall is pressed, called a first pinching surface, anda second pinching surface facing the first pinching surface. 4. The basket according to claim 3, wherein the end of the at least one wall is pinched in the housing, by being in mechanical contact with both lateral housing surfaces. 5. The basket according to claim 1, wherein the peripheral partition comprises an internal surface oriented inwardly of the basket and participating in delimiting the cells,wherein the housing opens to the internal surface, such that at least one of the lateral housing surfaces is orthogonal to the internal surface. 6. The basket according to claim 1, wherein a value ratio of a thickness of at least one wall to a surface length of wall lateral surface contact with one of the lateral housing surfaces, in a transverse cross-sectional plane of the basket, is between 0.2 and 1. 7. The basket according to claim 1, wherein the internal partition is in mechanical contact with the peripheral partition on at least ¾ the height of the peripheral partition. 8. The basket according to claim 7, wherein the internal partition is in mechanical contact with the peripheral partition over substantially the entire height of the basket. 9. The basket according to claim 1, comprising a plurality of internal partitions formed by stacked interlaced structural sets. 10. The basket according to claim 1, wherein the fastener comprises a plurality of tightening elements spaced from each other along the height of the basket. 11. The basket according to claim 10, wherein each fastener is configured to exert a tightening force the value of which is independent of that of the other tightening elements. 12. The basket according to claim 1, wherein the fastener is configured to be tightened/untightened from outside the basket. 13. The basket according to claim 1, wherein the fastener comprises a screw and a nut configured to cooperate with the screw. 14. The basket according to claim 13, wherein the fastener comprises at least one jaw, wherein the jaw is separation biased by the nut to press at least one of the lateral wall surfaces against at least one of the lateral housing surfaces. 15. The basket according to claim 1, wherein the fastener comprises at least one first tilted surface, the peripheral partition or the wall end comprising at least one second tilted surface complementary to the first tilted surface and configured to be supported on the first tilted surface. 16. The basket according to claim 2, wherein the fastener comprises at least one elastic tightening element located in the spacing, the elastic tightening element tending to press at least one of the lateral external surfaces against one of the lateral housing surfaces. 17. A transport and/or storage pack for radioactive materials, comprising a packaging and a lid closing the packaging, the packaging housing a basket according to claim 1. 18. The transport and/or storage pack according to claim 17, wherein the radioactive material comprises nuclear fuel assemblies. 19. A method for assembling a basket the basket, the basket comprising:at least one internal partition comprising at least one wall having two opposite lateral surfaces, andat least one peripheral partition,the internal partition delimits at least partially on either side of the same two cells intended to house the radioactive materials, the peripheral partition participating with the internal partition in delimiting the cells,the peripheral partition comprising at least one housing accommodating one end of the at least one wall, the housing comprising two opposite lateral housing surfaces and a bottom bringing together the two lateral housing surfaces,wherein the basket comprises a fastener configured to press at least one of the lateral wall surfaces against at least one of the lateral housing surfaces and to tighten the lateral wall surface against the lateral housing surface,the method comprising, after a step of accommodating the end of the wall in the housing, a step of pressing at least one of the lateral wall surfaces against at least one of the lateral housing surfaces, such that the wall is located between the fastener and the lateral housing surface against which the lateral wall surface is pressed. 20. The basket according to claim 1, wherein the fastener does not contact the at least one internal partition. |
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abstract | Initially, an ion beam is formed as an elongated shape incident on a wafer, where the shape has a length along a first axis longer than a diameter of the wafer, and a width along a second axis shorter than the diameter of the wafer. Then, a center of the wafer is moved along a scan path intersecting the ion beam at a movement velocity, and the wafer is rotated around at a rotation velocity simultaneously. During the simultaneous movement and rotation, the wafer is totally overlapped with the ion beam along the first axis when the wafer intersects with the ion beam, and the rotation velocity is at most a few times of the movement velocity. Both the movement velocity and the rotation velocity can be a constant or have a velocity profile relative to a position of the ion beam across the wafer. |
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summary | ||
abstract | A structure for preventing a scan by a beam is provided. The structure includes a primary material forming the structure. The primary material includes a first mass attenuation coefficient enabling the primary material to be penetrated by the beam. The structure also includes a matrix of dense particles within the primary material. The dense particles include secondary materials different than the primary material. The secondary materials comprise a subsequent mass attenuation coefficient that is greater than the first mass attenuation coefficient of the primary material. The subsequent mass attenuation coefficient enables the dense particles to attenuate the beam to distort the scan. |
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055330754 | description | BRIEF DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIGS. 2 and 3, a preferred embodiment of the alignment device of the invention is illustrated. FIGS. 2 and 3 also show a spent fuel container which basically corresponds to that discussed above in connection with FIG. 1 and elements of which are identified by the same reference numerals in FIGS. 2 and 3 with primes attached. The alignment device, which is generally denoted 20, includes a lightweight plate 22 which is fitted over the access opening 14' in the cover or lid 12' of the container 10'. Plate 22 is supported on cover 12' by a mounting arrangement including a plurality of pin holes 24 formed in plate 22 and a plurality of pins 26 disposed on the upper surface of cover 12'. It will be appreciated that the pins can be provided on plate 22 and the pinholes provided in cover 12' (as described below in connection with FIG. 4) and that other mounting arrangements can also be used. Plate 22 acts as a foundation or support for the installation and use of a laser 28. Thus, the main purpose of the mounting arrangement, i.e., the pins and pinholes of the exemplary embodiment being discussed, is to orient the laser 28 with respect to the centerline of the access opening 14' of the cover 12' of the spent fuel shipping container 10'. An O-ring 30 is installed on the underside of plate 22 around the periphery thereof. The plate 22 includes a viewing window 32 (not shown in FIG. 2) which enables visual observation of the laser beam for targeting purposes and a laser aperture or laser admittance window 34 (best seen in FIG. 2) disposed beneath laser 28 and through which the laser beam is transmitted. Plate 22 also preferably includes a pair of opposed handles (not shown) for permitting manual installation and removal thereof. Laser 28 preferably comprises a low power (approximately 0.5 mw) helium neon (HeNe) laser or visible diode laser. Such lasers are commercially available, self-contained and powered from batteries or a small 115 V a.c. adapter. An example of a suitable helium neon laser is that made by Melles Griot of Irvine, Calif. identified as Model 05 LLR 881 while an example of a suitable visible diode laser is the Model No. 06 DAL 001 (and 06 DAL 001/A) produced by the same manufacturer. A snap-in mount (not shown) is available as an accessory to enable fixed mounting of the laser device 28. The snap-in mount is attached to plate 22 with suitable fasteners (not shown). The mount for the laser 28 is shown schematically in the drawings and is generally denoted 36 in FIGS. 2 and 3. It is noted that a suitable mount is described in more detail in connection with FIG. 4. A suitable laser target 38 is incorporated into the top surface of each of the fuel pockets 16' of the spent fuel shipping container 10' during the container assembly process. The targets 38 are preferably reflective grooves or strips which provide enhanced reflection of the incoming beam. In the exemplary embodiment illustrated in FIGS. 2 and 3, the targets 38 are provided on the slanted shoulders 40 (FIG. 2) of the pockets 16'. A conventional indexing ring arrangement, represented by ring 42 (FIG. 3), is provided on the top of container 10' to indicate the proper angular position of cover 12' as is described in more detail below. In operation, prior to use of the system of the invention, the operability of the laser 28 is checked, as is the condition of the O-ring 30 on the bottom of plate 22. The laser 28 is mounted into the mount (indicated schematically at 36) and the perpendicularity of the laser 28 to the plate 22 is confirmed. A closure plug (not shown) is removed from the access opening 14' of the cover or lid 12' and the laser 28 is installed over the opening 14' and energized. The cover 12' of container 10' is then rotated and the laser beam is viewed through the viewing window 32 (see FIG. 3). Proper indexing is indicated when the reflection of the laser beam brightens because of the enhanced reflection provided by the target groove 28. The index ring arrangement 42 (FIG. 3) at the top of the container 10' permits identification of the azimuth reading which corresponds to the fuel pocket in question. Referring to FIG. 4, a further embodiment of the invention is shown. In FIG. 4, the cover is denoted 12" while elements similar to those of FIGS. 2 and 3 are given the same reference numerals with primes attached. In general, FIG. 4 simply illustrates some of the elements of FIGS. 2 and 3 in more detail, although there are also some differences. In FIG. 4, the support plate 22' is provided with pins or dowels (one of which, denoted 26', is shown) while the cover 12" includes the corresponding pinholes or recesses (one of which, denoted 24', is shown). More importantly, in accordance with this preferred embodiment, a single access window, denoted 32' is employed which provides both a laser mount and viewing access. As illustrated, window 32' is mounted in a recess 42' in plate 22' and secured to plate 22' by capscrews 44. Laser 28' is mounted by a mounting arrangement generally denoted 36' and including a support cage 46 and an L-shaped mounting bracket 48 secured to plate 22'. FIG. 4 also illustrates one of two gripping handles, denoted 50, which were mentioned above and which are located on opposite sides of plate 22'. It will be appreciated that the operation of the embodiment of FIG. 4 is basically the same as that of the embodiment of FIGS. 2 and 3. Although the present invention has been described relative to specific exemplary embodiments thereof, it will be understood by those skilled in the art that variations and modifications can be effected in these exemplary embodiments without departing from the scope and spirit of the invention. |
043326390 | summary | The present invention relates generally to nuclear reactors and more particularly to a failed element detection and location (FEDAL) system and method for use in a nuclear reactor and specifically an LMFBR. The particular nuclear reactor disclosed herein is one which utilizes a number of fuel assemblies housed within an active core which, in turn, is housed within a sealed vessel. Each fuel assembly contains the reactor's fuel or fission products, that is, the active substance making up the reactor, such as plutonium oxide. This active substance is sealed within a relatively large number of elongated hollow pins (cladding) located within an opened container having an inlet and outlet. Each of these containers and associated fuel pins comprise a single fuel assembly and all of the assemblies are located within the active core. The reactor also includes liquid metal cooling fluid such as liquid sodium and means for circulating a stream of the fluid along a path, a section of which passes through the containers from their inlets to their outlets. Obviously, this particular type of reactor includes other components, which may be conventional like those thus far described but which are not necessary to an understanding of the present invention. Accordingly, these other components will not be discussed or even mentioned herein, unless to do so would be helpful to an understanding of the present invention. In nuclear reactors of the type described, it is often desirable, if not necessary, to monitor for cladding failures, that is, breaks in the hollow fuel pins comprising part of the fuel assemblies. If this break is relatively small, the passing fluid, specifically the sodium, may not come in contact with the active substance within the pin, for example the plutonium oxide. However, inert gases including specifically Kr-85, Kr-88 and Xe-133, Xe-135 will escape into the fluid stream, emitting gamma rays therefrom. On the other hand, if the break is relatively large, that is, sufficiently large to cause the passing sodium to actually enter the faulty pin and contact the plutonium oxide, the sodium will be contaminated with I-137 and Br-87 which are two of a number of by-products of the fissioning process taking place in the pin and which decay rather rapidly, giving off neutrons. The detection of small breaks is relatively conventional and typically accomplished by detecting for gamma rays emitted from the escaping gases Kr-85, 88 and the like as the latter surface from a central pool of sodium within the reactor vessel. However, accurate and reliable detection of the larger breaks in a reliable manner is not as simple, as will be seen hereinafter. One typical way of monitoring for large breaks heretofore has been to place a neutron detector at some entry point in the internal heat exchanger (IHX) which also comprises part of the overall reactor and which is located within the reactor vessel for receiving liquid sodium after the latter passes out of the fuel assembly and into the central pool. There are several problems with this approach. First, it may not be possible to locate the monitoring apparatus in a position to collect samples of sodium which have passed through all of the core assemblies because of the size of the IHX and the diverse ways in which the sodium enters the latter. Second, the time it takes for the sodium to reach the IHX bulk sodium pickup point within the central pool is relatively long. This means that by the time these contaminants are detected for the emission of neutrons, the level of neutrons being emitted will be relatively low thereby raising the question of reliability. For example, the half life for I-137 is approximately 55 seconds and for Br-87 it is approximately 22 seconds. In contrast to this, it may take as long as 150 seconds for a particular sample of sodium to reach the selected entry point of the IHX from the reactor core. Another way in which the relatively larger breaks in fuel pins have been detected in the past has been to individually sample each fuel assembly, one at a time, which can certainly be reliable. However, it is time consuming and costly to provide continuous individual monitoring of all of the fuel assemblies since a given core may be made up of as many as 600-700 such assemblies. As will be seem hereinafter the present invention provides for a particular FEDAL approach for use in a nuclear reactor of the type described without the previously recited drawbacks. Rather, as will also be seen, combined samples of liquid sodium are collected as soon as the latter passes through selected groups of fuel assembly containers while the neutron emission level of any collected contaminants is still relatively high, thereby making this approach reliable. Should there be an indication of a break, individual samples are then and only then taken to isolate the fuel assembly or assemblies responsible for the break. In this way, individual samples do not have to be continuously collected and detected as in the past. In view of the foregoing, one object of the present invention is to provide a reliable and yet economical FEDAL technique for use in a nuclear reactor of the type described above. A more particular object of the present invention is to provide a reliable technique of detecting for relatively large breaks in the previously described fuel pins wherein combined samples of sodium are collected at the outlets of selected fuel pin containers while the neutron emission level of the contaminants, if any, are still relatively high. Another particular object of the present invention is to collect individual sodium samples only if the combined sample indicates a break in one or more fuel pins. A further object of the present invention is to provide a FEDAL system which utilizes an uncomplicated and reliable valve assembly for collecting both combined sodium samples as well as individual samples. As stated previously and as will be seen hereinafter, particular FEDAL technique disclosed herein is one which is especially suitable for use in a particular type of nuclear reactor, specifically an LMFBR. As also stated, this type of reactor has a reactive core housing within a vessel and a plurality of fuel assemblies housed within the core. Each of these fuel assemblies includes an open container having an inlet and outlet and active substance such as plutonium oxide sealed within a relatively large number of elongated hollow pins located within the container. This reactor also includes liquid metal cooling fluid such as liquid sodium and means for circulating a stream of the fluid along a path, a section of which passes through the containers from their inlets to their outlets. As will be seen hereinafter, the particular technique disclosed is one which detects breaks in the hollow fuel pins of sufficient size to cause at least one predetermined contaminant to pass into the liquid metal cooling stream as the latter passes through the containers. In accordance with this technique, a combined sample of the liquid metal fluid is collected at the outlets of at least a group of the fuel assembly containers and detected for the presence or absence of the contaminant. If this combined sample indicates the presence of a break, individual samples of the fluid are selectively collected, one at a time, at the outlets of the fuel assembly containers in the same group and these individual samples are also detected for the presence or absence of the contaminant, thereby indicating the particular fuel assembly or assemblies responsible for the break. |
055442072 | claims | 1. An apparatus for measuring the thickness of a nonmagnetic overlay clad of a ferromagnetic pressure vessel of a nuclear reactor comprising: a magnetic yoke; one or more exciting coils, said coils wrapped around said magnetic yoke for magnetizing said magnetic yoke to form a magnetic path through said magnetic yoke and the ferromagnetic pressure vessel of the nuclear reactor; means for measuring the spatial distribution of the magnetic field component orthogonally crossing the ferromagnetic pressure vessel of the nuclear reactor. 2. An apparatus is in claim 1, wherein said means for measuring the spatial distribution of the magnetic field component orthogonally crossing the ferromagnetic pressure vessel of the nuclear reactor is a plurality of magnetic field sensors placed in a line along the longitudinal direction of said magnetic yoke above the nonmagnetic overlay clad. 3. An apparatus is in claim 1, wherein said means for measuring the spatial distribution of the magnetic field component orthogonally crossing the ferromagnetic pressure vessel of the nuclear reactor is a single magnetic field sensor moveable along the longitudinal direction of said magnetic yoke above the nonmagnetic overlay clad. |
041692292 | summary | BACKGROUND OF THE INVENTION The invention relates to an apparatus for keying in electron beams, appropriate for a raster electron scan microscope, an electron beam recording device, and the like. Such apparatus comprises a deflection system for the deflection of the electron beam and controllable as a function of time for deflecting the beam to effect a scanning operation. A velocity modulating effect adapted to the beam velocity is used for controlling deflection. The use of an electron beam keying apparatus is known from the state of the art in an electron beam system of a raster electron scan microscope. The actual apparatus for keying in the electron beam is a deflection system by which the electron beam is deflected in the regular case between the cathode and the raster deflection field over an aperture lens and/or it is deflected into and out of said aperture lens. Such a deflection system preferably is mounted behind the first anode, that is behind the principal electron beam (viewed from the cathode). Thereby it is at least proximal to the area of the beam focusing. Moreover, it is a customary practice to carry out with a raster electron scan microscope potential contrast measurements, particularly at integrated circuits of the semiconductor type. It is possible to record with these readings potential courses electron-optically, for example in conductor tracks, whereby a particular interest exists in potential changes at very high frequencies and impulse currents of high frequencies. Because the evaluation of the secondary electron emission utilized with potential contrast measurements with electron beams as such is at least inappropriate for recording very rapid processes, for example with 10.sup.7 to 10.sup.9 cycles, true in time, the so-called stroboscope principle has been applied which is known from high frequency measuring techniques under the term "sampling". Thereby a repeated correlated scanning of a desired spot takes place, whose potential contrast is to be determined. This stroboscopic scanning then is repeated at a slight phase shifting, so that, in line with the type of a sampling method the entire, very high-frequency potential course can be determined, although in the measuring direction a relative time carrier effect is involved, namely in connection with the secondary emission and the collection of the secondary electrons. This stroboscope or sampling method requires the production of extremely short electron beam impulses in the raster electron scan microscope, whose phase then is to be shifted moreover, still in comparison with the AC signal in the sample to be examined. This phase shifting is tantamount with a shifting of the impulse moment of the electron beam. Such short electron beam impulses are realized according to known practical operation with the use of an aperture lens and such a beam deflection apparatus which basically is provided with two elements, placed opposite each other and extending in the axial direction of an electron beam, and together forming a deflection capacitor, whereby, however, a deflector plate is opposed by an installation which is a traveling wave installation with interdigital structure. In said interdigital structure the wave of the deflecting signal with which this impulse scanning deflecting installation is to be impinged moves at the same speed as the beam electrons passing through this deflecting capacitor. An interdigital structure in the meaning of the invention relates to a finger conduit to be impinged with the signal and interdigitated with a second finger conduit which is grounded. Additional details can be obtained from the printed publications IEEE Transactions on Electron Devices, volume ED-19, No. 2 (1972), page 204-213 and Scanning Electron Microscopy (1973) (part I), Proc. of the Sixth Annual Electron Microscope Symposium, IIT Research Institute, Chicago, Ill. 60616, USA (April 1973) and Scanning Electron Microscopy (1976) (Part IV) Proc. of the Workshop on Microelectronic Device Fabrication and Quality Control with SEM, IIT Research Institute (April 1976). Such a meander conduit with interdigital structure is shown particularly in FIG. 1 of the article appearing in the first-named printed publication IEEE Transactions on Electron Devices. These cited references also show the manner of operation of a known raster scan electron microscope and/or of the deflection system with traveling wave included in the electron beam system of such a microscope. One disadvantage of a deflection system known from such prior art consists in that the transverse deflection field in relation to the electron beam axis prevailing between the meander structure and the opposite electrode necessarily also has a longitudinal electrical field component which thus causes an acceleration or retardation of the electron beam. This leads to a defocusing of the electron beam focused as such extremely sharply on the sample to be tested, such defocusing occuring in operation as the electric deflection signal is varied to produce deflection. SUMMARY OF THE INVENTION This defocusing effect recognized as a basis of the present invention in devices according to the state of the art, shall be eliminated as a problem in accordance with the teachings of the present invention. According to the invention this problem is solved in an apparatus as recited in the preamble of claim 1, by the features recited in the characterizing part of claim 1, and further improvements and perfections of the invention are recited in the dependent claims. The solution according to the invention is based on the idea of providing in the deflection system used as a basis, such measures which will insure that the longitudinal component of the transverse deflection field considered as disadvantageous and undesirable as such will be eliminated. According to the solution offered under the invention this is accomplished by the fact that the deflection installation is of such a symmetrical design that the two deflection parts contained in the deflection system as "deflection plates" are identical with each other and placed symmetrically in relation to each other with reference to the electron beam axis. This measure according to the invention accomplishes that with the supply, likewise according to the invention, of deflection signals with opposite signs, these relations cause at these deflection parts that in the electron beam axis a uniform axial potential prevails, even when the electron beam deflection process starts and is carried out. Under the invention the electron beam continues to be focused sharply upon deflection out of its central position in the opening of the apertured lens. Additional explanations of the invention become evident from the following description of an embodiment; and other objects, features and advantages will be apparent from this detailed disclosure and from the appended claims. |
abstract | Disclosed is a method for fabricating fine conductive patterns using a surface modified mask template, the method including: depositing a high molecular substance on a substrate; applying a hydrophobic material onto the high molecular substance so that the hydrophobic material can infiltrate into the high molecular substance; forming a mask template by removing a part of the high molecular substance to form a recess where a region of the substrate is exposed to an outside; depositing conductive ink on the mask template; and performing annealing to abstract metal particles from a metallic compound dissolved in the conductive ink so that an insulating pattern can be formed in a region on which the high molecular substance is deposited, but a conductive pattern can be formed as the metal particles are abstracted from the conductive ink in the recess and cohere with each other. |
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048329053 | description | PREFERRED EMBODIMENTS OF THE INVENTION The numeral 10 generally designates a fuel assembly unit. The fuel assembly 10 includes an upper end fitting 12, guide tubes 14, spacer grids 16 supporting fuel rods 17, and a skirt portion 18 shown partially broken away in FIG. 1 to illustrate a lower end fitting debris collector constructed primarily of Inconel according to the principles of the invention, generally designated by the numeral 20. Behind the skirt 18 are debris trapping compartments defined by the lower end fitting debris collector grid 21 and solid fuel rod end caps 22. Each fuel rod end cap 22 is located at the intersection of first intersecting and slottedly interlocked grid forming bars 24, known as top bars because of their lower slots 27. Second grid forming bars 26, which are known as bottom bars because of their upper slots 27 for slottedly interlocking with the grid forming bars 24 along their upper margin. When assembled, the bars 24 in the area of intersection are above the bars 26. The pairs of first and second intersecting and slottedly interlocked grid forming bars 24 and 26 are attached to the perimeter strip 28. Members 24, 26 and 28 may be of Inconcel 718. The upper bars or thick strips 24 are parallel to each other and the lower bars or thick strips 26 are parallel to each other and normal to the upper bars 24. All of the bars 24 and 26 are welded or brazed to the rectangular perimeter strip 28. The grid 21 is completed by thin Inconel or other suitable metal strips 30 which are parallel to and mounted between the bars 24 and 26, typically, in groups of two to four parallel strips 30. The strips 30 mounted in slots 32 of bars 24 are normal to the strips 30 mounted in slots 32 of bars 26. As shown, the strips 30 mounted in slots 32 of bars 24 are in engagement with the strips 30 mounted in slots 32 of bars 26, but this is not necessary. All of the strips 30 are welded or brazed to the perimeter strip 28. It has been found advantageous in welding to lay a weld bead 34 along the length of the upper edges of bars 26 to hold the bars 24 and 30 of the top side of the grid 21 in assembled relation therewith and to lay a weld bead 36 along the length of the lower edges of bars 24 to hold the bars 26 and 30 of the bottom side of the grid 21 in assembled relation therewith, as shown in FIG. 3. A lower end fitting debris collector base member 40 is provided which is, typically, a casting of a suitable stainless steel alloy such as A.I.S.I. 304. The base member 40 is welded to bars 24 and 26. In the form illustrated, the base 40 has hollow legs 42 with fastener receiving interior cavities or bores 44 and fastener seats 46. The legs 42 are joined by web members 48 and each leg is welded or brazed into an opening cut into or formed in the grid 21. A flat upper surface 48 of each leg 42 is basically flush with the grid 21 and supports a guide tube 14. A fastener 50 with a slotted head seated against fastener seat 46 is threaded into the end of an internally threaded end of guide tube 14 to secure the guide tube 14 and lower end fitting debris collector 20. Typically, four guide tubes 14 are thus secured. In the central portion of grid 21, an insert 51 is provided instead of a leg 42 for providing means to fasten the central guide tube 14. Insert 51 is hollow and has a fastener seat 46a for retaining the head of a fastener 50. It is welded or brazed within an opening in grid 21. As seen in the embodiment illustrated by schematic FIG. 7, where modified grid parts have like numerals to those of the embodiment of FIGS. 1-6 but with the subscript "a", the intersections of bars 24a and 26a form contoured seats 60 for fuel rod end caps 22. Otherwise, the end caps merely sit on the intersecting strips 24 and 26. The contoured seats for fuel rod end caps 22 illustrated in FIG. 7, are formed by bevels 62 at the open ends of slots 27a of bars 26a. The bars 24a have notches 64 with the same taper as bevels 62 and at locations along the upper edges of bars 24a at the points of intersection with bars 26a. Striss 30 are not affected by seats 60. In all filtering concepts, it must be noted that filterings will not be 100%. Certain deleterious debris can still escape. The effectiveness of the filter is determined by the amount of debris it will filter. To determine filtering effectiveness, a series of tests of proof of principal wer conducted to measure other devices' filtering ability as compared to the instant invention. The filtering effectiveness of the novel modified lower end fitting of the invention proved to be better than most other devices with the advantage of not affecting fuel rod plenum volume, fuel rod expansion shoulder gap, fuel rod reconstitutability and fuel rod support. Therefore, while it is not possible to entirely eliminate fuel rod failure due to debris, the risk of failures can be greatly reduced by means of the instant invention, thereby increasing fuel assembly reliability without compromising design integrity in the abovementioned ways. |
description | The present invention relates to a radiation shielding method and a radiation shielding device applied when an operation such as a plant outage or repair is performed in a nuclear power plant, for example. In nuclear power plants, plant outages are performed for the structure thereof. Appropriate repair is performed for a part where it is considered that repair is required according to the plant outage. Thus, in nuclear power plants, operations such as plant outages and repair are required to maintain normal operating conditions. In such operations, it is necessary to reduce an amount of radiation to workers. Taking this necessity into account, there can be considered an installation of a wall-shaped shielding material that shields radiation in a structure, which is an object to be shielded. However, to reduce the amount of radiation, a shielding material having a weight of, for example, 100 kilograms or more is required. It is not easy to transport such a heavy shielding material to a checking location or a repair location. Further, there is an idea that the shielding material can be divided into pieces of about 10 kilograms; however, because a long installation time is required in a higher or narrower location, there is a concern about exposure to radiation of workers at the time of installing the shielding material. Conventionally, for example, Patent Literature 1 discloses a pipe cleaning method in which a cleaning area and a non-cleaning area of a pipe are isolated from each other by simple means. According to this cleaning method, a balloon is inserted into a boundary between the cleaning area and the non-cleaning area of the pipe, air or fluid such as water is supplied into the balloon to pressurize the balloon, thereby isolating the cleaning area of the pipe from the non-cleaning area. That is, it can be considered to apply the conventional cleaning method at the time of performing a plant outage or repair of a nuclear power plant, in which radiation is easily shielded by a shielding body in which water is filled in a balloon, thereby reducing the amount of radiation to a worker. Patent Literature 1: Japanese Patent Application Laid-open No. 2003-80192 However, according to the radiation shielding method and the radiation shielding device that supplies water into a balloon, although it is effective in a place where the amount of radiation is relatively small, radiation may not be shielded effectively in a place where the amount of radiation is relatively large. The present invention has been achieved to solve the problems described above, and an object of the present invention is to provide a radiation shielding method and a radiation shielding device that can reduce an amount of radiation to a worker easily and sufficiently. According to an aspect of the present invention, a radiation shielding method includes: installing a hollow container at a predetermined portion of an object to be shielded; feeding fluid into the container via a feeding hose; and supplying a shielding material to the feeding hose and transporting and filling a granular shielding material into the container by the fluid. According to the radiation shielding method, a worker approaches the object to be shielded at the time of installing the container. However, because the granular shielding material is fed together with fluid into the container installed in the object to be shielded at a remote place via the feeding hose, the worker does not need to approach the object to be shielded, and further, the shielding effect can be improved by the granular shielding material. Therefore, the amount of radiation to the worker can be reduced easily and sufficiently. Advantageously, in the radiation shielding method, at the feeding fluid into the container, liquid is used as the fluid and the liquid is filled in the container via the feeding hose. According to the radiation shielding method, because the shielding material settles down in the fluid filled in the container and gradually accumulates on a bottom of the container, the shielding material can be filled in the container tidily, and a radiation shielding effect can be obtained sufficiently. Advantageously, the radiation shielding method further includes: extracting the shielding material filled in the container from the container together with fluid discharged to outside of the container via a returning hose, while feeding fluid into the container via a feeding hose, in a state that the shielding material is filled in the container; and recovering the shielding material from the fluid. According to the radiation shielding method, because the shielding material can be recovered from the container together with fluid at a remote place from the object to be shielded, a worker does not need to approach the object to be shielded, thereby enabling to reduce the amount of radiation to the worker easily and sufficiently. Advantageously, in the radiation shielding method, the container is mounted on the object to be shielded at all times. According to the radiation shielding method, an operation of installing the container in the object to be shielded can be omitted at the time of a plant outage or repair, thereby enabling to further reduce the amount of radiation to the worker. According to another aspect of the present invention, a radiation shielding device includes: a hollow container installed at a predetermined portion of an object to be shielded; a fluid feeding unit that feeds fluid into the container via a feeding hose; and a shielding-material supply unit that supplies a granular shielding material to the feeding hose. According to the radiation shielding device, the radiation shielding method described above can be performed. As a result, a worker approaches the object to be shielded at the time of installing the container. However, because the granular shielding material is fed together with fluid into the container installed in the object to be shielded at a remote place via the feeding hose, the worker does not need to approach the object to be shielded, and further, the shielding effect can be improved by the granular shielding material. Therefore, the amount of radiation to the worker can be reduced easily and sufficiently. Advantageously, the radiation shielding device includes: a shielding-material extracting unit that circulates the shielding material filled in the container together with fluid discharged to outside of the container via a returning hose, while feeding fluid into the container via a feeding hose; and a shielding-material recovering unit that recovers the shielding material from the fluid. According to the radiation shielding device, the radiation shielding method described above can be performed. As a result, because the shielding material can be recovered from the container together with fluid at a remote place from the object to be shielded, a worker does not need to approach the object to be shielded, thereby enabling to reduce the amount of radiation to the worker easily and sufficiently. Advantageously, in the radiation shielding device, the shielding-material extracting unit includes an injection nozzle that injects the fluid fed into the container, and a fetching member having an inlet for fetching the shielding material together with fluid discharged from the container, which are provided in the container, and an injection port of the injection nozzle is arranged toward the inlet of the fetching member. According to the radiation shielding device, because fluid is injected from the injection port of the injection nozzle toward the inlet of the fetching member, a swirling current is generated at a position of the inlet. Therefore, the shielding material near the inlet is stirred by the swirling current and introduced into the fetching member from the inlet, and extracted to the returning hose. As a result, clogging of the shielding material at the inlet can be avoided. Advantageously, in the radiation shielding device, the shielding-material extracting unit includes a switching unit that switches a feeding direction of fluid in a reverse flow mode of the fluid. According to the radiation shielding device, by feeding fluid in a reverse direction by the switching unit, the fluid flowing back in the returning hose for discharging the fluid to outside of the container is fed into the container. Therefore, the shielding material is blown into the container, thereby removing clogging. Advantageously, in the radiation shielding device, the hose for circulating the shielding material together with fluid between the shielding-material supply unit and the container is made to be transparent. According to the radiation shielding device, the shielding material being fed via the hose can be visually checked, and thus clogging of the shielding material can be recognized. Advantageously, in the radiation shielding device, the hose for circulating the shielding material together with fluid between the shielding-material recovering unit and the container is made to be transparent. According to the radiation shielding device, the shielding material being fed via the hose can be visually checked, and thus clogging of the shielding material can be recognized. Advantageously, in the radiation shielding device, water is used as the fluid, and a pellet containing tungsten is used as the shielding material. According to the radiation shielding device, a pellet containing tungsten obtained by solidifying tungsten powder in a granular form by a resin material can be reused for subsequent radiation shielding, and also can be incinerated. As a result, handling of the pellet used for radiation shielding becomes easy. According to the present invention, because a granular shielding material is fed together with fluid into a container installed in an object to be shielded at a remote place via a hose, a worker does not need to approach the object to be shielded, and the shielding effect can be improved by the granular shielding material. Therefore, the amount of radiation to the worker can be reduced easily and sufficiently. Exemplary embodiments of a radiation shielding method and a radiation shielding device according to the present invention will be explained below in detail with reference to the accompanying drawings. The present invention is not limited to the embodiments. In addition, constituent elements in the following embodiments include those that can be easily replaced by persons skilled in the art or that are substantially equivalent. The radiation shielding method and the radiation shielding device according to the present invention are applied to general nuclear power plants such as a pressurized water reactor (PWR) and a boiling water reactor (BWR). Particularly, the radiation shielding method and the radiation shielding device according to the present invention are suitable when operations such as a plant outage and repair are performed in the general nuclear power plants. FIGS. 1 and 2 are schematic diagrams of a radiation shielding device according to an embodiment of the present invention, and FIG. 3 is a schematic diagram of an injection nozzle and a fetching member of a shielding-material extracting unit. As shown in FIG. 1, the radiation shielding device according to the present embodiment includes a container 1 installed at a predetermined portion of an object to be shielded 100, a fluid feeding unit 2 that feeds fluid into the container 1, and a shielding-material supply unit 3 that supplies a granular shielding material to the fluid fed into the container 1. The container 1 shown in FIG. 1 is formed in a shape that covers a periphery of the object to be shielded 100 in a tubular shape, in order to reduce exposure to radiation from the object to be shielded 100 in a tubular shape. The container 1 is formed in a hollow shape, and for example, made of a material having flexibility and retractility such as stainless steel, plastic, or urethane rubber. When the container 1 is made of stainless steel or plastic, a high rigidity can be obtained. Meanwhile, when the container 1 is made of urethane rubber, because it can be folded small due to its flexibility, it is suitable for transport, and the container 1 can be closely stuck together with the object to be shielded 100 due to its retractility. Although not shown, it is preferable that an observation window is formed in the container so that the condition thereof can be visually checked from outside. The fluid feeding unit 2 includes a tank 21, a hose 22, and a pump 23. Fluid to be fed into the container 1 is stored in the tank 21. For the fluid, water, pure water, boric-acid solution, polyvinyl alcohol, or silicon oil is used as a liquid, and air is used as a gas. In the present embodiment, the tank 21 is shown as a storage for storing liquid. The hose 22 connects the container 1 with the tank 21 to feed fluid between the container 1 and the tank 21, and includes a feeding hose 22a for feeding fluid from the tank 21 to the container 1, and a returning hose 22b for returning fluid from the container 1 to the tank 21. The hoses 22a and 22b are respectively connected to connection ports 1a and 1b provided in an upper part of the container 1. At least the feeding hose 22a of the hose 22 is made to be transparent, so that internal flowage can be visually checked from outside. The pump 23 is disposed intermediate of the feeding hose 22a to pump fluid in the tank 21 to the container 1. The fluid feeding unit 2 feeds fluid stored in the tank 21 into the container 1 via the feeding hose 22a by an operation of the pump 23, and returns the fluid filled in the container 1 to the tank 21 via the returning hose 22b. The shielding-material supply unit 3 is constituted as a so-called hopper that stores a shielding material and causes a fixed quantity of the shielding material to drop from a funnel-shaped bottom port of a drop-bottom type. The shielding-material supply unit 3 is provided on a downstream side of the pump 23 provided in the feeding hose 22a in the fluid feeding unit 2. As the shielding material stored in the shielding-material supply unit 3, pellets containing tungsten obtained by solidifying tungsten powder in a granular form by a resin material, stainless steel grains obtained by processing stainless steel in a granular form, lead grains obtained by processing lead in a granular form, and depleted uranium grains obtained by processing depleted uranium 1 in a granular form are used. It is preferable that such a shielding material is formed in the same grain shape and the same grain size so that deposition in the container 1 is equalized. The shielding-material supply unit 3 supplies the stored shielding material to the feeding hose 22a. The supplied shielding material is pumped together with fluid in the feeding hose 22a and filled in the container 1. The shielding material in an amount to be filled in the container 1 is stored in the hopper as the shielding-material supply unit 3. Although not shown, a filter is provided in the connection port 1b of the container 1 connected with the returning hose 22b, so that the grains of the shielding material are not returned to the tank 21 together with the fluid. Although not shown, the pump 23 of the fluid feeding unit 2 and the shielding-material supply unit 3 are mounted together on a carriage so that transport can be facilitated. As shown in FIG. 2, the radiation shielding device according to the present embodiment further includes a shielding-material extracting unit 4 that extracts the shielding material from the container 1 together with fluid discharged from the container 1, while feeding fluid into the container 1, and a shielding-material recovering unit 5 that recovers the shielding material from fluid. The shielding-material extracting unit 4 includes a tank 41, a hose 42, and a pump 43. Fluid to be fed into the container 1 is stored in the tank 41. The tank 21 of the fluid feeding unit 2 can be also used as the tank 41. For the fluid, water, polyvinyl alcohol, or silicon oil is used as a liquid, and air is used as a gas. In the present embodiment, the tank 41 is shown as a storage for storing liquid. The hose 42 connects the container 1 with the tank 41 to feed fluid between the container 1 and the tank 41, and includes a feeding hose 42a for feeding fluid from the tank 41 to the container 1, and a returning hose 42b for returning fluid from the container 1 to the tank 41. The hoses 42a and 42b are respectively connected to connection ports 1c and 1d provided on a bottom of the container 1. At least the returning hose 42b of the hose 42 is made to be transparent, so that internal flowage can be visually checked from outside. The pump 43 is disposed intermediate of the feeding hose 42a to pump fluid in the tank 41 to the container 1. The shielding-material extracting unit 4 includes an injection nozzle 44 and a fetching member 45 provided on the bottom of the container 1. As shown in FIG. 3, the injection nozzle 44 is formed in a tubular shape, with one end thereof communicating with the connection port 1c to which the feeding hose 42a is connected, and the other end being closed. A plurality of injection ports 44a are provided in the injection nozzle 44 along an extending direction in a tubular shape. The fetching member 45 is formed in a tubular shape, with one end thereof communicating with the connection port 1d to which the returning hose 42b is connected, and the other end being closed. A plurality of inlets 45a is provided in the fetching member 45 along an extending direction in a tubular shape. The fetching member 45 is arranged on the bottom of the container 1, with the inlets 45a being directed upward. The injection nozzle 44 is arranged alongside the fetching member 45, with the injection ports 44a being directed toward the inlets 45a of the fetching member 45. In the present embodiment, the injection nozzle 44 is arranged above the fetching member 45, with the injection ports 44a being directed downward. The shielding-material extracting unit 4 has a switching unit 46 for the feeding hose 42a and the returning hose 42b. The switching unit 46 includes first and second bypass pipes 46a and 46b that connect the feeding hose 42a and the returning hose 42b to each other. The first and second bypass pipes 46a and 46b cross each other and are connected to the feeding hose 42a and the returning hose 42b. The switching unit 46 also includes switching valves 46c, 46d, 46e, and 46f. The switching valve 46c is arranged between positions in the feeding hose 42a where the first and second bypass pipes 46a and 46b are respectively connected thereto, to allow flowage of fluid in an opened state, while stopping flowage of fluid in a closed state. The switching valve 46d is arranged between positions in the returning hose 42b where the first and second bypass pipes 46a and 46b are respectively connected thereto, to allow flowage of fluid in an opened state, while stopping flowage of fluid in a closed state. The switching valve 46e is arranged in the first bypass pipe 46a, to allow flowage of fluid in an opened state, while stopping flowage of fluid in a closed state. The switching valve 46f is arranged on the second bypass pipe 46b, to allow flowage of fluid in an opened state, while stopping flowage of fluid in a closed state. The shielding-material extracting unit 4 feeds fluid stored in the tank 41 into the container 1 via the feeding hose 42a by an operation of the pump 43, with the switching valves 46c and 46d of the switching unit 46 being opened, and the switching valves 46e and 46f being closed (a direct flow mode). The fluid filled in the container 1 is then returned to the tank 41 via the returning hose 42b. When the fluid is returned from the container 1 to the tank 41 via the returning hose 42b, the shielding material in the container 1 is circulated together with the fluid into the returning hose 42b. The fluid fed into the container 1 via the feeding hose 42a is injected, as shown in FIG. 3, from the injection ports 44a of the injection nozzle 44 toward the inlets 45a of the fetching member 45, thereby causing a swirling current at the positions of the inlets 45a. Therefore, a shielding material D near the inlets 45a is introduced into the pipe of the fetching member 45 from the inlets 45a together with the fluid, while being stirred by the swirling current, and extracted to the returning hose 42b. The shielding-material extracting unit 4 feeds fluid stored in the tank 41 into the container 1 via the returning hose 42b, bordering on the switching unit 46, as shown by the arrow of one-dot-chain line, by the operation of the pump 43, with the switching valves 46c and 46d of the switching unit 46 being closed, and the switching valves 46e and 46f being opened (a reverse flow mode). The fluid filled in the container 1 is then returned to the tank 41 via the feeding hose 42a. In this manner, when the fluid is reversely fed, the fluid is fed into the container 1 from the inlets 45a of the fetching member 45. Therefore, the shielding material D near the inlets 45a is blown into the container 1. The shielding-material recovering unit 5 stores the shielding material. The shielding-material recovering unit 5 is provided in the returning hose 42b between the switching unit 46 and the tank 41 in the shielding-material extracting unit 4. Further, the shielding-material recovering unit 5 is connected to the returning hose 42b via a filter 5a. The filter 5a causes the fluid fed by the returning hose 42b to flow directly, while stopping and dropping the shielding material into the shielding-material recovering unit 5. Although not shown, the pump 43 and the switching unit 46 of the shielding-material extracting unit 4, and the shielding-material recovering unit 5 are both mounted on a carriage so that transport can be facilitated. According to the radiation shielding method using the radiation shielding device configured in this manner, the hollow container 1 is first installed at a predetermined portion of the object to be shielded 100. The fluid feeding unit 2 and the shielding-material supply unit 3 are then installed. At this time, the connection ports 1c and 1d of the container 1 are closed. Fluid (liquid is used here as the fluid) is fed to the container 1 via the fluid feeding unit 2, thereby filling the container 1 with the fluid. The shielding material is supplied by the shielding-material supply unit 3, while feeding the fluid into the container 1 by the fluid feeding unit 2. Accordingly, the shielding material is fed into the container 1. At this time, the shielding material settles down in the fluid filled in the container and gradually accumulates on the bottom of the container. Further, because the feeding hose 22a is made to be transparent, the shielding material being fed through the feeding hose 22a can be visually checked, thereby enabling to recognize clogging of the shielding material in the feeding hose 22a. Further, if an observation window is formed in the container 1, an internal condition in which the shielding material accumulates can be visually checked and recognized. When the shielding material is filled in the container 1, feed of fluid by the fluid feeding unit 2 is suspended, to remove the fluid feeding unit 2 and the shielding-material supply unit 3 and close the connection ports 1a and 1b of the container 1. As a result, the shielding material is filled in the container together with fluid, and thus exposure to radiation from the object to be shielded 100 can be reduced. When shielding of radiation is not required, the container 1 is removed from the object to be shielded 100, as described below. First, the shielding-material extracting unit 4 and the shielding-material recovering unit 5 are installed. The switching unit 46 is turned into the reverse flow mode, to feed fluid into the container 1 by the shielding-material extracting unit 4. Accordingly, because fluid is fed from the inlets 45a of the fetching member 45 into the container 1, the shielding material near the inlets 45a is blown into the container 1, thereby removing clogging at the inlets 45a. The switching unit 46 is then turned to the direct flow mode, to feed fluid into the container 1 by the shielding-material extracting unit 4. The fluid filled in the container 1 is fed to the returning hose 42b together with the shielding material. The fluid is returned to the tank 41 by the shielding-material recovering unit 5, while the shielding material is stored in the shielding-material recovering unit 5. Accordingly, the shielding material filled in the container 1 is stored in the shielding-material recovering unit 5. Further, because the returning hose 42b is made to be transparent, the shielding material fed through the returning hose 42b can be visually checked, thereby enabling to recognize clogging of the shielding material in the fetching member 45 and the returning hose 42b. When there is clogging of the shielding material in the fetching member 45 or the returning hose 42b, the switching unit 46 is turned to the reverse flow mode to feed the fluid to the container 1 by the shielding-material extracting unit 4, thereby feeding the shielding material together with fluid from the inlets 45a of the fetching member 45 into the container 1 to remove clogging of the shielding material. When the entire shielding material filled in the container 1 is stored in the shielding-material recovering unit 5, feed of fluid by the shielding-material extracting unit 4 is suspended, and the shielding-material extracting unit 4, the shielding-material recovering unit 5, and the container 1 are removed, to finish the operation. The radiation shielding method according to the present embodiment includes a step of installing the hollow container 1 at a predetermined portion of the object to be shielded 100, a step of feeding fluid into the container 1 via the feeding hose 22a, and a step of supplying the shielding material to the feeding hose 22a to transport and fill a granular shielding material into the container by the fluid. According to the radiation shielding method, a worker approaches the object to be shielded 100 at the time of installing the container 1 and the hose 22 of the fluid feeding unit 2. However, in other cases, because the granular shielding material is fed into the container 1 together with fluid via the feeding hose 22a at a remote place from the object to be shielded 100, a worker does not need to approach the object to be shielded 100. Further, because the shielding effect can be improved by the granular shielding material, the amount of radiation to the worker can be reduced easily and sufficiently. In the radiation shielding method according to the present embodiment, it is preferable that liquid is used as the fluid and filled in the container 1 via the feeding hose 22a at the step of feeding the fluid into the container 1 via the feeding hose 22a. According to the radiation shielding method, because the shielding material settles down in the fluid filled in the container and gradually accumulates on the bottom of the container, the shielding material can be tidily filled in the container 1, thereby enabling to obtain the sufficient shielding effect of radiation. The radiation shielding method according to the present embodiment further includes a step of extracting the shielding material filled in the container 1 from the container together with the fluid discharged to the outside of the container 1 via the returning hose 42b, while feeding the fluid into the container 1 via the feeding hose 42a in a state that the shielding material is filled in the container 1, and a step of recovering the extracted shielding material. According to the radiation shielding method, because the shielding material can be recovered from the container 1 together with the fluid at a remote place from the object to be shielded 100, a worker does not need to approach the object to be shielded 100, thereby enabling to reduce the amount of radiation to the worker easily and sufficiently. Further, in the radiation shielding method according to the present embodiment, it is preferable to mount the container 1 on the object to be shielded 100 at all times. According to the radiation shielding method, an operation of installing the container 1 on the object to be shielded 100 can be omitted at the time of a plant outage or repair, thereby enabling to further reduce the amount of radiation to the worker. The radiation shielding device according to the present embodiment described above includes the hollow container 1 installed at a predetermined portion of the object to be shielded 100, the fluid feeding unit 2 that feeds fluid into the container 1 via the feeding hose 22a, and the shielding-material supply unit 3 that supplies a granular shielding material to the feeding hose 22a. According to the radiation shielding device, the radiation shielding method described above can be performed. As a result, a worker approaches the object to be shielded 100 at the time of installing the container 1 and the hose 22 of the fluid feeding unit 2. However, in other cases, because the shielding material is fed to the container 1 together with the fluid at a remote place from the object to be shielded 100, the worker does not need to approach the object to be shielded 100. Further, because the shielding effect can be improved by the granular shielding material, the amount of radiation to the worker can be reduced easily and sufficiently. The radiation shielding device according to the present embodiment includes the shielding-material extracting unit 4 that circulates the shielding material together with the fluid discharged to the outside of the container 1 via the returning hose 42b, while feeding the fluid into the container 1 via the feeding hose 42a, and the shielding-material recovering unit 5 that recovers the shielding material from the fluid. According to the radiation shielding device, the radiation shielding method described above can be performed. As a result, because the shielding material can be recovered from the container 1 together with the fluid at a remote place from the object to be shielded 100, a worker does not need to approach the object to be shielded 100, thereby enabling to reduce the amount of radiation to the worker easily and sufficiently. Further, in the radiation shielding device according to the present embodiment, the shielding-material extracting unit 4 includes, in the container 1, the injection nozzle 44 that injects fluid fed into the container 1, and the fetching member 45 having the inlets 45a for fetching the shielding material together with the fluid discharged from the container 1, and the injection ports 44a of the injection nozzle 44 are arranged towards the inlets 45a of the fetching member 45. According to the radiation shielding device, because fluid is injected from the injection ports 44a of the injection nozzle 44 toward the inlets 45a of the fetching member 45, a swirling current is generated at positions of the inlets 45a. Therefore, the shielding material near the inlets 45a is introduced into the pipe of the fetching member 45 from the inlets 45a, while being stirred by the swirling current, and is extracted to the returning hose 42b. As a result, clogging of the shielding material at the inlets 45a can be avoided. Particularly, in the radiation shielding device according to the present embodiment, the feeding hose 42a of the shielding-material extracting unit 4 is connected to the connection port 1c provided on the bottom of the container 1, and the shielding material is extracted from the feeding hose 42a into the container 1 together with fluid. Therefore, the shielding material accumulating on the bottom of the container 1 can be appropriately extracted. In the radiation shielding device according to the present embodiment, the shielding-material extracting unit 4 includes the switching unit 46 that switches a feeding direction of fluid in a mode in which the fluid is reversely fed. According to the radiation shielding device, by reversely feeding fluid by the switching unit 46, the fluid is fed into the container 1 from the inlets 45a of the fetching member 45. Therefore, the shielding material near the inlets 45a is blown into the container 1, thereby removing clogging at the inlets 45a. In the radiation shielding device according to the present embodiment, the feeding hose 22a and the returning hose 42b that circulate the shielding material together with fluid are made to be transparent. According to the radiation shielding device, the shielding material being fed via the feeding hose 22a and the returning hose 42b can be visually checked, and thus clogging of the shielding material can be recognized. In the radiation shielding device according to the present embodiment, water is used as the fluid, and a pellet containing tungsten is used as the shielding material. According to the radiation shielding device, water and the pellet containing tungsten can be reused for subsequent radiation shielding, and also can be incinerated. As a result, handling of what has been used for radiation shielding is facilitated. In the radiation shielding device according to the present embodiment, the feeding hose 22a of the fluid feeding unit 2 is connected to the connection port 1a provided in the upper part of the container 1, and the shielding material supplied from the feeding hose 22a by the shielding-material supply unit 3 is fed into the container 1 together with fluid. Therefore, because the shielding material reaches the bottom of the container 1 from above, the shielding material can accumulate appropriately in the container 1. Further, in the radiation shielding method according to the present embodiment, after fluid (liquid is used here as the fluid) is filled in the container 1, the shielding material is supplied together with the fluid. Therefore, because the shielding material settles down in the liquid filled in the container 1 and gradually accumulates on the bottom of the container 1, the shielding material can accumulate appropriately in the container 1. FIGS. 4 to 8 are schematic diagrams of a container used in the radiation shielding device. A container 11 shown in FIG. 4 is applied when the object to be shielded 100 is a valve installed in a pipe. In this case, it is preferable to use a pair of containers 11, 11 in combination, of which shapes are matched with the valve shape so that the valve is sandwiched from both sides. The respective containers 11, 11 are integrated by male and female engaging members 7 and fitted to the valve. Each of the containers 11 is provided with the connection ports 1a, 1b, 1c, and 1d like in the container 1, and although not shown, the injection nozzle 44 and the fetching member 45 are provided in the container 11 like in the container 1. A container 12 shown in FIG. 5 is applied when the object to be shielded 100 is a pipe. In this case, it is preferable to use a pair of containers 12, 12 in combination, of which shapes are matched with the shape of the pipe so that the pipe is sandwiched from both sides. The respective containers 12, 12 are integrated by the male and female engaging members 7 and fitted to the pipe. Each of the containers 12 is provided with the connection ports 1a, 1b, 1c, and 1d like in the container 1, and although not shown, the injection nozzle 44 and the fetching member 45 are provided in the container 12 like in the container 1. A container 13 shown in FIG. 6 is applied when the object to be shielded 100 is a large tank. In this case, it is preferable to use a plurality of wall-like containers 13, 13, 13, 13, and 13 to enclose the circumference of the tank. The respective containers 13, 13, 13, 13, and 13 are integrated by male and female engaging members and fitted to the circumference of the tank, although not shown. Each of the containers 13 is provided with the connection ports 1a, 1b, 1c, and 1d like in the container 1, and although not shown, the injection nozzle 44 and the fetching member 45 are provided in the container 13 like in the container 1. A container 14 shown in FIG. 7 and a container 15 shown in FIG. 8 are applied to a maintenance work of a steam generator nozzle in a nuclear power plant. For example, as a maintenance work of an inlet nozzle 103 of an inlet-side water chamber 102 of a steam generator 101, when repair of a welded part 106 between an elbow pipe 105 that connects the inlet nozzle 103 with a primary cooling pipe 104 and the inlet nozzle 103 is to be performed, inner walls of the inlet-side water chamber 102 and the primary cooling pipe 104 are the objects to be shielded 100. In repair of the welded part 106, because a worker enters into the inlet-side water chamber 102 from a manhole 102a, the container 14 is installed to follow the inner wall of the inlet-side water chamber 102 (see FIG. 7), and the container 15 is installed to block the inside of the primary cooling pipe 104 (see FIG. 8). Installation of the container 14 shown in FIG. 7 is performed according to procedures shown in FIGS. 9 to 12. A support member 8 for supporting the container 14 is used here. The support member 8 forms a frame constituted of a stainless steel pipe material arranged to cover an opening 103a of the inlet nozzle 103 inside the inlet-side water chamber 102, and defines a desired work area around the opening 103a of the inlet nozzle 103. The support member 8 includes enclosing parts 8a in a downward U-shape arranged in parallel, extending across the opening 103a of the inlet nozzle 103, and a connecting part 8b that connects upper parts of the enclosing part 8a. The enclosing part 8a and the connecting part 8b are divided into a plurality of numbers, and brought into the inlet-side water chamber 102 from the manhole 102a by a worker. Further, to install the support member 8 inside the inlet-side water chamber 102, a base unit 9 is arranged on the bottom of the inlet-side water chamber 102. The base unit 9 is fitted to the bottom of the inlet-side water chamber 102 and laid therein, as shown in FIGS. 9 and 10, with the opening 103a of the inlet nozzle 103 and the manhole 102a being opened. Mounting holes 9a are formed on the base unit 9, into which respective ends of the enclosing parts 8a of the support member 8 are inserted. The base unit 9 is constituted by a member having a strength sufficient for supporting the support member 8 inserted into the mounting holes 9a, such as an aluminum plate, and includes a member that shields radiation, for example, a shielding material in which a plurality of tungsten sheets formed by mixing tungsten powder with a resin material are stacked on each other. The base unit 9 is divided into a plurality of numbers so that these divided parts are brought into the inlet-side water chamber 102 from the manhole 102a by a worker. When the base unit 9 does not include the mounting holes 9a, the base unit 9 is constituted only by the tungsten sheets. The container 14 forms a so-called balloon in which a shell made of urethane rubber or the like and having flexibility and retractility is covered by high frequency welding in a pouch-like shape so that the inside becomes hollow. The container 14 is put between the support member 8 installed inside the inlet-side water chamber 102 and an inner wall 100 of the inlet-side water chamber 102, and is divided into a plurality of parts. In the present embodiment, in FIGS. 7, 11, and 12 depicting a mode in which fluid is filled therein and the shell is inflated, the container 14 is divided into a first container 14a arranged in an inner region of the inlet-side water chamber 102 farthest from the manhole 102a (see FIGS. 7 and 11), a second container 14b arranged in a circular-arc side region of the inlet-side water chamber 102 (see FIGS. 11 and 12), a third container 14c arranged in a side region on a partition board 102b side of the inlet-side water chamber 102 (see FIGS. 11 and 12), and a fourth container 14d arranged in an upper region of the inlet-side water chamber 102 (see FIGS. 7, 11, and 12). A partition wall (not shown) that divides the hollow part into a plurality of rooms is provided in a container that shields a relatively large region such as the fourth container 14d, so that an inflated shape does not deform. The partition wall is made of a material same as that of the shell (urethane rubber or the like), and has a plurality of holes so that respective rooms communicate with each other. Although not shown, the containers 14 (14a, 14b, 14c, and 14d) are provided with the connection ports 1a, 1b, 1c, and 1d like in the container 1, and the injection nozzle 44 and the fetching member 45 are provided in the containers 14 (14a, 14b, 14c, and 14d) like in the container 1. To install the containers 14 (14a, 14b, 14c, and 14d) inside the inlet-side water chamber 102, after the support member 8 is installed inside the inlet-side water chamber 102, the deflated first container 14a is arranged at a predetermined position between the support member 8 and the inner wall 100 of the inlet-side water chamber 102 and air is supplied thereto to inflate the first container 14a. The deflated second container 14b is arranged at a predetermined position between the support member 8 and the inner wall 100 of the inlet-side water chamber 102 and air is supplied thereto to inflate the second container 14b. The deflated third container 14c is arranged at a predetermined position between the support member 8 and the inner wall 100 of the inlet-side water chamber 102 and air is supplied thereto to inflate the third container 14c. Next, water is supplied to the first container 14a, the second container 14b, and the third container 14c in this order to replace air by water, and the shielding material is filled therein. The deflated fourth container 14d is then arranged at a predetermined position between the support member 8 and the inner wall 100 of the inlet-side water chamber 102 and air is supplied thereto to inflate the fourth container 14d, followed by supply of water to replace air by water, and the shielding material is filled therein. The containers 14 (14a, 14b, 14c, and 14d) filled with the shielding material in this manner are combined in the inlet-side water chamber 102, to cover the opening 103a of the inlet nozzle 103 as a work area. Because the amount of radiation to the worker from the inner wall 100 of the inlet-side water chamber 102 is reduced by the containers (14a, 14b, 14c, and 14d) filled with the shielding material, the operation can be performed safely. Images of the condition of the containers 14 (14a, 14b, 14c, and 14d) can be taken by a camera and monitored by a monitor outside of a structure. On the other hand, the container 15 shown in FIG. 8 forms a so-called balloon in which a shell made of urethane rubber or the like and having flexibility and retractility is covered by high frequency welding in a pouch-like shape so that the inside becomes hollow. Although not shown, the container 15 is provided with the connection ports 1a, 1b, 1c, and 1d like in the container 1, and the injection nozzle 44 and the fetching member 45 are provided in the container 15 like in the container 1. Installation of the container 15 is performed according to procedures shown in FIGS. 8, and 13 to 16. First, the container 15 in which the shell is in a deflated mode, the feeding hose 22a is connected to the connection port 1a, and the returning hose 22b is connected to the connection port 1b is brought into the inlet-side water chamber 102 from the manhole 102a by a worker, and the container 15 is caused to slide from the inlet nozzle 103 into the primary cooling pipe 104 through the elbow pipe 105. When the position and orientation of the container 15, which is caused to slide into the primary cooling pipe 104, are not appropriate, the worker adjusts the position and orientation of the container 15 from the inlet-side water chamber 102, by using a guide member 10 (see FIG. 13). The guide member 10 is a long stick and the length thereof can be adjusted by expanding and contracting the guide member 10. An upward U-shaped hook 10a and a lock pin 10b that opens and closes an opening of the hook 10a are provided at a tip of the guide member 10. The lock pin 10b is opened and closed on a base side of the guide member 10, which is held by a worker. The hook 10a is hooked on a locking part 15a provided in the container 15 in a state that the lock pin 10b is opened, and then the hook 10a is locked on the locking part 15a in a state that the lock pin 10b is closed. Therefore, the position and orientation of the container 15 can be adjusted without any need of the worker to enter into the primary cooling pipe 104 (see FIGS. 14(a) and 14(b)). Further, when the container 15 with the shell being deflated is arranged in the primary cooling pipe 104, respective holding members 15b provided on both sides of the container 15 come in contact with an inner bottom face of the primary cooling pipe 104. That is, the holding members 15b form legs for arranging the container 15 inside the primary cooling pipe 104. Therefore, the container 15 before the shell is inflated can be maintained in the position and orientation adjusted inside the primary cooling pipe 104. The worker then brings a camera C into the inlet-side water chamber 102 from the manhole 102a, and installs the camera C at a position where the container 15 can be checked from the inlet nozzle 103. The worker then exits the inlet-side water chamber 102, so that there is nobody in the structure. Accordingly, in the structure in an unmanned state, images of the condition of the container 15 are taken by the camera C. Images taken by the camera C are monitored by a monitor outside the structure (see FIG. 15). Air is then supplied into the container 15. During sir supply, when it is confirmed from the images on the monitor that the position and orientation of the container 15 have changed due to inflation of the shell, a worker enters into the inlet-side water chamber 102 from the manhole 102a, to adjust the position and orientation of the container 15 by the guide member 10. In this manner, air is filled in the container 15, while monitoring the condition of the container 15 by the monitor outside the structure and appropriately adjusting the position and orientation of the container 15. Thereafter, water is supplied into the container 15 to replace air by water, and the shielding material is filled therein (see FIG. 16). In this manner, the container 15 filled with the shielding material blocks the primary cooling pipe 104, while coming in contact with the inner wall 100 of the primary cooling pipe 104. Because the amount of radiation to a worker irradiated from the primary cooling pipe 104 toward the inlet-side water chamber 102 is reduced by the container 15 filled with the shielding material, the operation can be performed safely. The container 15 is removed according to procedures shown in FIGS. 17 and 18. First, a worker enters into the inlet-side water chamber 102 from the manhole 102a to connect the feeding hose 42a to the connection port 1c of the container, and connect the returning hose 42b to the connection port 1d. The worker then extracts and recovers the shielding material from the container 15. Thereafter, the worker pulls up the container 15 with the shell being deflated from the primary cooling pipe 104 to the inlet-side water chamber 102 by the guide member 10 (see FIG. 17). As shown in FIG. 18, an L-shaped hook 10c is provided at the tip of the guide member 10. On the other hand, a pull-up rope 15c is provided on the container 15, and a loop is formed at the end of the pull-up rope 15c. As shown in FIGS. 17 and 18, by hooking the hook 10c of the guide member 10 into the loop of the pull-up rope 15c and pulling it up, the pull-up rope 15c comes to hand of the worker. By holding the loop of the pull-up rope 15c and pulling the pull-up rope 15c, the worker can pull the deflated container 15 up to the inlet-side water chamber 102 (see FIG. 17). Finally, the container 15 is brought out to outside of the inlet-side water chamber 102 from the manhole 102a. The container 15 is removed in this manner. As explained above, in the radiation shielding device according to the present embodiment, by applying various containers such as the containers 1, 11, 12, 13, 14, 15, it is possible to perform radiation shielding of various parts. Industrial Applicability As described above, the radiation shielding method and the radiation shielding device according to the present invention are suitable for easily and sufficiently reducing an amount of radiation to a worker. Reference Signs List 1, 11, 12, 13, 14 (14a, 14b, 14c, 14d), 15 container 1a, 1b, 1c, 1d connection port 2 fluid feeding unit 21 tank 22 hose 22a feeding hose 22b returning hose 23 pump 3 shielding-material supply unit 4 shielding-material extracting unit 41 tank 42 hose 42a feeding hose 42b returning hose 43 pump 44 injection nozzle 44a injection port 45 fetching member 45a inlet 46 switching unit 46a first bypass pipe 46b second bypass pipe 46c, 46d, 46e, 46f switching valve 5 shielding-material recovering unit 5a filter 7 engaging member 100 object to be shielded (inner wall) C camera D shielding material |
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055240313 | abstract | A method for retrofitting an irradiated nuclear fuel assembly including removing a portion of the inlet nozzle and installing a debris filter. |
claims | 1. An autonomous self-powered system for cooling radioactive materials, the system comprising:a pool at least partially filled with a liquid and radioactive materials immersed in the liquid;a closed-loop fluid circuit comprising a working fluid having a boiling temperature that is less than a boiling temperature of the liquid, the closed-loop fluid circuit comprising, in an operable fluid coupling, an evaporative heat exchanger at least partially immersed in the liquid, a turbogenerator, and a condenser;one or more forced flow units operably coupled to the closed-loop fluid circuit to induce flow of the working fluid through the closed-loop fluid circuit; andthe closed-loop fluid circuit converting thermal energy extracted from the liquid of the pool into electrical energy that powers the one or more forced flow units;wherein the evaporative heat exchanger comprises:a top header, a bottom header, a downcomer tube defining a first passageway between the top and bottom headers, and a plurality of heat exchange tubes each forming a second passageway between the top and bottom headers, each of the plurality of heat exchange tubes comprising an outer surface, and wherein an entirety of the outer surface of the heat exchange tubes are in direct contact with the liquid of the pool;a working fluid inlet extending into the downcomer tube for introducing a liquid phase of the working fluid into the first passageway; anda working fluid outlet for allowing a vapor phase of the working fluid to exit the evaporative heat exchanger. 2. The autonomous self-powered system of claim 1 wherein the downcomer tube and the heat exchange tubes of the evaporative heat exchanger are arranged in a substantially vertical orientation and the downcomer tube is circumferentially surrounded by the heat exchange tubes. 3. The autonomous self-powered system of claim 1 wherein the downcomer tube comprises a thermal insulating layer. 4. The autonomous self-powered system of claim 1 wherein the first passageway has a first width and the each of the second passageways have a second width, the first width being greater than the second width. 5. The autonomous self-powered system of claim 1 wherein the downcomer tube has a first effective coefficient of thermal conductivity and the heat exchange tubes have a second effective coefficient of thermal conductivity, the first effective coefficient of thermal conductivity being less than the second effective coefficient of thermal conductivity. 6. The autonomous self-powered system of claim 1 wherein the evaporative heat exchanger is configured to achieve an internal thermosiphon flow of the liquid phase of the working fluid within the evaporative heat exchanger. 7. The autonomous self-powered system of claim 1 further comprising:the evaporative heat exchanger converting the working fluid from the liquid phase to the vapor phase by transferring the thermal energy from the liquid in the pool to the working fluid;the turbogenerator receiving the vapor phase of the working fluid from the evaporative heat exchanger, the turbogenerator generating the electrical energy by extracting energy from the vapor phase of the working fluid flowing through the turbogenerator;the condenser receiving the vapor phase of the working fluid from the turbogenerator and converting the vapor phase of the working fluid flowing through the condenser back into the liquid phase of the working fluid by removing thermal energy from the working fluid; andthe at least one forced flow unit electrically coupled to the turbogenerator so as to be powered by the electrical energy generated by the turbogenerator. 8. The autonomous self-powered system of claim 1 wherein the one or more forced flow units comprises a hydraulic pump, the closed-loop fluid circuit comprising the hydraulic pump, the hydraulic pump forcing the liquid phase of the working fluid into the evaporative heat exchanger. 9. The autonomous self-powered system of claim 1 wherein the liquid in the pool is at a first pressure and the working fluid within the evaporative heat exchanger is at a second pressure that is greater than the first pressure, the boiling temperature of the working fluid at the second pressure being less than the boiling temperature of the liquid in the pool at the first pressure. 10. The autonomous self-powered system of claim 1 wherein the liquid in the pool is water and the working fluid is a refrigerant or a hydrocarbon. 11. The autonomous self-powered system of claim 1 wherein the evaporative heat exchanger is fully immersed in the liquid in the pool and located at a top portion of the pool. 12. The autonomous self-powered system of claim 1 wherein the autonomous self-powered system operates free of electrical energy generated outside of the closed-loop fluid circuit. 13. A vertical evaporative heat exchanger for immersion in a heated fluid comprising:a tubeside fluid circuit comprising:a top header;a bottom header;a core tube forming a downcomer passageway between the top header and the bottom header, the core tube having a first effective coefficient of thermal conductivity;a plurality of heat exchange tubes forming passageways between the bottom header and the top header, the plurality of the heat exchange tubes having a second effective coefficient of thermal conductivity that is greater than the first effective coefficient of thermal conductivity;a working fluid in the tubeside fluid circuit;a conduit extending through the top header of the tubeside fluid circuit, the conduit comprising:a working fluid inlet in a first end of the conduit for introducing a liquid phase of the working fluid into the tubeside fluid circuit; anda working fluid outlet in a second end of the conduit that is located within the downcomer passageway so that the working fluid flows directly into the downcomer passageway;at least one vapor outlet for allowing a vapor phase of the working fluid to exit the top header; andwherein transfer of heat from the heated fluid to the working fluid induces a thermosiphon flow of the liquid phase of the working fluid within the tubeside fluid circuit. 14. The vertical evaporative heat exchanger of claim 13 further comprising a plurality of the vapor outlets, and wherein the working fluid inlet of the conduit is surrounded by the vapor outlets. 15. The vertical evaporative heat exchanger of claim 14 further comprising a longitudinal axis, and wherein the conduit extends along the longitudinal axis. 16. The autonomous self-powered system of claim 1 wherein the evaporative heat exchanger further comprises a conduit extending through the top header of the evaporative heat exchanger and into the downcomer tube, the conduit comprising the working fluid inlet in a first end of the conduit and a working fluid outlet in a second end of the conduit, the second end of the conduit being located within the downcomer tube to ensure that the liquid phase of the working fluid is introduced directly into the downcomer tube to isolate the working fluid being introduced into the evaporative heat exchanger via the conduit from the working fluid already circulating within the evaporative heat exchanger. 17. The autonomous self-powered system of claim 1 wherein the downcomer tube extends from a first end to a second end and each of the heat exchange tubes extends from a first end to a second end, and wherein the first end of the downcomer tube is aligned with the first end of each of the heat exchange tubes and wherein the second end of the downcomer tube is aligned with the second end of each of the heat exchange tubes. 18. The autonomous self-powered system of claim 17 wherein the top header comprises a top tube sheet and the bottom header comprises a bottom tube sheet, the first end of each of the heat exchange tubes and the downcomer tube terminating at the top tube sheet and the second end of each of the heat exchange tubes and the downcomer tube terminating at the bottom tube sheet. 19. The autonomous self-powered system of claim 1 wherein the heat exchange tubes are not enclosed within a housing such that the outer surface of outermost ones of the heat exchange tubes form a portion of an outer surface of the evaporative heat exchanger. 20. The autonomous self-powered system of claim 1 wherein each of the heat exchange tubes extends from a first end that is connected to the top header to a second end that is connected to the bottom header, the heat exchange tubes having a length measured from the first end to the second and, and wherein the outer surfaces of the heat exchange tubes are in direct contact with the liquid in the pool along an entirety of the length of the heat exchange tubes. |
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abstract | A high-polymer neutron shielding material which scarcely reduces the hydrogen number density when exposed to a high temperature of 150 to 200xc2x0 C. for a long time period |
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049884735 | claims | 1. An arrangement for reducing the reactivity of a spent nuclear fuel assembly to permit placement thereof in a fuel storage facility, said arrangement comprising: (a) an elongated tool; (b) a reactivity-reducing device including (c) means separate from said elongated tool and said reactivity-reducing device and being detachably attachable to said trailing end of said rod for closing said passage therethrough and rendering said latch means at said opposite leading end of said rod inaccessible by said elongated tool from said trailing end of said rod such that said latch means remains at said latching position, said elongated tool being insertable through said passage of said rod upon detaching of said closing means from said rod for engaging and moving said latch means against said biasing means from said latching position to said releasing position. 2. The arrangement as recited in claim 1, wherein said latch means includes a plurality of latch members being mounted for pivotal movement between displaced latching and releasing positions at the leading end of said rod. 3. The arrangement as recited in claim 2, wherein said means also includes a plurality of actuating levers respectively attached to said latch members and engageable by said elongated tool when inserted through said passage for moving said levers and thereby said latch members from their latching to releasing positions once said closing means has been detached from said rod trailing end. 4. The arrangement as recited in claim 1, wherein said closing means is a closure plug threadably attached to said trailing end of said rod. 5. The arrangement as recited in claim 1, wherein said rod has a tubular portion of a diameter smaller than an inside diameter of a guide thimble for permitting insertion of said tubular portion therein and a head portion attached to said tubular portion at said trailing end of said rod and of a diameter larger than the inside diameter of the guide thimble such that said head portion is capable of suspending said tubular portion in the guide thimble when said rod is inserted therein. 6. The arrangement as recited in claim 5, wherein said head portion has means defined thereon for cooperating with an independent remover for gripping said head portion to withdrawal said rod from the guide thimble once said self-latching means have been unlatched from the guide thimble. |
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summary | ||
042833671 | summary | The present invention relates to a method of separating krypton from radioactive waste gases which become free during the chemical dissolution of burned-off core fuel particles and which contain krypton and xenon, and also concerns a gas separation plant for practicing said method. When regenerating fuel elements, the core fuel particles are chemically dissolved in order to separate the fission products or the disintegration products thereof, which form during the reactor operation, from the fuel and/or breeder elements usable again for making fuel elements. A portion of the fission products is formed by the gases krypton and xenon which are contained in the waste gas of the chemical dissolver. To blow off the waste gas into the atmosphere is, due to the radioactive isotope Kr.sup.85 contained in the krypton and comprising a halfway time of 10.3 years, not tolerable in an unlimited manner. Therefore, it is necessary to store the waste gases occurring during the chemical dissolution of the core fuel particles, until the radioactivity has dropped to a minimum which is sufficiently safe. For storing the waste gases, considerable space is required. The krypton component in the waste gas is, however, relatively small. With reference to the xenon volume, the krypton volume is only approximately 15/85. When depositing the entire waste gas which is generated during the chemical dissolution of the core fuel particles, the major portion of the storage spaces have to be made available for such gas components the nature of which is not dangerous for the surroundings. This space requirement is rather uneceonomical and represents a considerable disadvantage. It is, therefore, an object of the present invention to provide a method of and device for separating krypton from waste gases which become free during the chemical dissolution of core fuel particles to be generated, and which method and device will permit a continuous removal of the krypton gas from the waste gas mixture. |
abstract | A method for tomographic nuclear imaging determines the distance between a non-parallel hole collimator surface and a region of interest (ROI) by obtaining difference images between images acquired at different view angles of the ROI. The distance may be used in a nuclear image reconstruction algorithm to more accurately reconstruct an image of the ROI. The method takes advantage of the non-stationary Point Spread Function of a non-parallel hole collimator to determine depth information of gamma events emitted from the ROI. |
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062051962 | description | DESCRIPTION OF EMBODIMENTS OF THE INVENTION Embodiments of the present invention will be explained hereunder, referring to the drawings. FIG. 1 shows an embodiment of the present invention. In this embodiment, in a boiling water type nuclear reactor in which a plurality of fuel assemblies 2 each surrounded by a channel box 1 are loaded, and a plurality of control rods having control rod blades each disposed between the channel boxes, long blade control rods 6 each having control rod blades extending latintudinal in four directions, respectively, are arranged between the channel boxes on a diagonal of each of square bundle regions each formed of a plurality of (four in this embodiment) the fuel assemblies 2, and short blade control rods 7 each are arranged between channel boxes of each of the square bundle regions at the center of the region, each of which short blade control rods 7 has a blade length (in a lateral or latitudinal direction) of about one half of the width of one of the square bundle regions, for example, substantially the same as the width of each of the above-mentioned fuel assemblies. With this construction, as mentioned above, in the long blade control rod 6 arranged between the channel boxes on the diagonal, the control rod worth per one rod increases and the number of control rods and the number of control rod driving devices can be reduced by the number corresponding to an increment of the control rod worth, so that the cost can be reduced. Quantitatively, the number of the control rods can be reduced by 25% as compared with the conventional lattice. FIG. 2 shows an arrangement of the control rods over the whole reactor core. Symbols .omicron. denote the latitudinal long blade control rods 6 and symbols .circle-solid. denote the latitudinal short blade control rods 7. Moreover, as clearly shown in both FIGS. 1 and 2, the long blade control rods each have a blade length in a latitudinal direction which is about twice as long the blade length in a latitudinal directional of the short blade control rods. It is found that the number of control rods and the number of the control rod driving devices can be reduced largely as compared with the conventional arrangement and the control rod system can be simplified. Further, as explained previously in the summary of the invention, the reactor shutdown margin can be secured easily and the number of control rods is reduced. As a result, Gd for securing a reactor shutdown margin does not remain and low inventory fuel is not loaded, whereby economy is improved greatly. Further, in this embodiment, by sharing the role of the control rods such that the long blade control rods on the diagonal serve for reactor shutdown and the short blade control rods at central portions are for controlling reactivity during operation and at time of scram, the system can be rationalized and simplified, and the cost of the whole plant can be reduced. Further, in the short blade control rods for controlling reactivity, by using a neutron absorber of material (B.sup.10) which has a high reactivity effect, the control rod worth of the short blade control rods increases, and scram characteristic and reactivity control characteristic can be increased. Further, since the long blade control rods on the diagonal are not used for scram, a control system of high speed scram, etc. can be omitted, which enables use of a hydraulic driving system of a low cost, whereby a cost is reduced largely. Further, in the above-mentioned embodiment, it is possible to share the role of the control rods such that the long blade control rods 6 on the diagonal are used for controlling reactivity during operation and for shutdown of the reactor and the central short blade control rods 7 are used for scram. In this case, the system is rationalized and simplified as mentioned above, so that reduction of the cost can be expected. Further, in the arrangement as shown in FIG. 1, another embodiment, in which the reactivity worth of a control rod is improved at an upper region thereof, is explained hereunder with respect to a neutron absorber used in a control rod. In this embodiment, in particular, enrichment of B.sup.10 in the short blade control rod arranged at the central portion of the square bundle is made relatively high at the upper region. In general, in a boiling water type nuclear reactor, since a void ratio is higher at an upper region of the reactor during operation, neutron spectrum is hardened, and production of Pu.sup.239 by neutron absorption is promoted. Therefore, the enrichment of fissionable materials becomes high at an upper portion of the reactor and the reactor shutdown margin in the region decreases relatively. In this embodiment, the enrichment of B.sup.10 in the upper region of the length of the control rod is increased for the upper region of the nuclear core in which a reactor shutdown margin decreases relatively, whereby the reactor shutdown margin can be increased, as shown in FIG. 4 which is a graph showing the ratio of the high B.sup.10 enriched region of the control rod to the reactor shutdown margin. Further, since an amount of used B.sup.10 can be reduced, a manufacturing cost can be reduced. Therefore, a cost of the whole plant can be reduced in total. FIGS. 5 and 6 show a conventional core in part and a core in part according to the present invention, each of which is adopted for fuel assemblies of fuel rod lattice structure of 9.times.9. FIG. 6 shows an embodiment of the present invention in which a large-sized fuel assembly is formed by 4 mini-bundles each of which has a bundle width of about 12 inches (30.5 cm) as used in BWR and ABWR at present. As in the previous embodiments, in this embodiment in FIG. 6, a cost of the plant can be reduced largely by reduction of the number of control rods. Further, by role sharing of the control rods, a control rod system can be simplified, and a cost the whole plant can be reduced largely. Although not illustrated, fuel rod lattice structures of 8.times.8 and 10.times.10 also can be applied. FIGS. 7A and 7B show another embodiment of the present invention. In this embodiment, fuel assemblies 2 constituting a square bundle region as shown in FIG. 1 each have nine (9) water rods 11 as shown in FIG. 7A. As shown in FIG. 7B, each water rod 11 has an ascending flow path 12 and a descending flow path 13, the ascending and descending flow paths 12, 13 are connected to an inflow hole 14 and an outflow hole 15, respectively, and the inflow hole 14 is positioned at a portion lower than the outflow hole 15. The density of water in each water rod in this embodiment changes largely according to a flow rate of water passing through the fuel assembly. That is, under the condition that a flow rate of water in the core is small, since an amount of steam generated in the water rod becomes larger than an amount of water flowing in the control rod, the water rod inside is filled with steam. When the flow rate of water increases, an amount of water flowing in the water rod goes beyond an amount of steam generated therein, so that the water rod inside is filled with water. Therefore, the water rod inside is filled with steam in operating at a low flow rate in an initial burning stage, whereby an average density of water inside the fuel assembly decreases, so that neutron spectrum is hardened, whereby production of Pu.sup.239 is promoted. On the other hand, since the water rod inside is filled with water in operation at a high flow rate in a final burning stage and the average density of water inside the fuel assembly increases, the neutron spectrum is and it is possible to effectively burn Pu.sup.239 produced in the operation at a low flow rate, and fuel economy is raised. That is, since excessive neutrons in the initial burning stage can be used for production of Pu.sup.239, the number of short blade control rods can be reduced by a decrease in reactivity control by absorption of excessive neutrons during operation, so that further cost reduction is realized. Further, since a lot of the water rods can be arranged by making the fuel assembly large in size, an effect of effective use of Pu.sup.239 increases and fuel economy can be improved greatly. In this invention, since the fuel assemblies each are made large in size as the fuel assemblies in FIG. 1, it is effective from a viewpoint of fuel economy to provide, inside each fuel assembly, water rods the cross-sectional area of each of which corresponds to that of several fuel rods. Further, the reactor shutdown margin can be improved by using such material that control rod worth becomes high at a portion facing central side portions of the fuel assembly 2, as a neutron absorber arranged inside the control rod blades. According to the present invention, the number of control rods can be drastically reduced without decreasing control rod worth. Further, since the role of the control rod can be shared, the control system can be simplified and rationalized and a cost of the plant can be reduced. |
summary | ||
abstract | A filter capable of adjusting spectrum in multiple stages and that capable of attaining the reduction of size, as well as an X-ray imaging system having such a filter, are provided. The filter, which is for adjusting the spectrum of passing radiation, comprises a support plate having an aperture for passage therethrough of radiation, plural filter plates supported by the filter plate and having mutually different filter characteristics, and moving device for moving the plural filter plates selectively to a position to close the aperture and a position to open the aperture. |
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042886996 | claims | 1. A storage rack for the storage of fuel elements of nuclear reactors consisting of a sheet metal lattice arrangement constituting a plurality of abutting similar vertical storage cases having a rectangular cross section, characterized in that said lattice arrangement is constructed of a plurality of mutually substantially perpendicular sheet elements having a sharply shaped contour, the sheet elements in one of the two perpendicular directions being continuous traversing sheet elements forming walls and extending over the widths of a plurality of cases of the lattice arrangement, two adjacent traversing sheet elements being spaced at a mutual distance substantially corresponding with the width of a storage case and all sheet elements in the other of the two perpendicular directions being narrow sheet elements each extending only substantially over the width of one storage case, the narrow sheet elements at the opposite long edges thereof being provided with lugs protruding from the edges, the traversing sheet elements having openings formed therein for receiving the lugs, said lugs having openings extending therethrough for receiving wedges, the opening in the lug being positioned on the opposite side of a traversing sheet element from the narrow sheet element when the lug is inserted through an opening in the traversing sheet element, and wedges insertable into the openings in the lugs so that the long edges of the narrow sheet elements may be pressed tightly against the traversing sheet elements. 2. The storage rack of claim 1, characterized in that the lugs of narrow sheet elements between the two juxtaposed traversing sheet elements are offset with respect to the lugs of narrow sheet elements in the adjoining regions between the traversing sheet elements. 3. The storage rack of claim 1 or 2, characterized in that within the lattice arrangement all storage columns have been separated from each other by a wall consisting of two parallel sheet elements at a short distance from each other with interstices formed between these sheet elements and that the lugs and the wedges have been situated within these interstices. 4. The storage rack of claim 1, characterized in that two narrow sheet elements together constitute a partition wall, the lugs on the elements being provided at the same level, a connecting strip running parallel to the traversing sheet elements for connecting lugs to each other and a key pinched between the connecting strip and the opposite wall of the cooperating traversing sheet element. 5. The storage rack of claim 4, characterized in that the connecting strips are welded to the lugs. 6. The storage rack of claim 4, characterized in that upon assembling the wedges are welded to the traversing sheet elements, the connecting strips and the lugs, respectively. 7. The storage rack of claim 1, characterized in that at each of the four outer walls of the rectangular lattice arrangement the storage cases are enclosed by a traversing sheet element extending over the entire side wall length and that the sheet elements of the side walls extending parallel to the narrow sheet elements are provided with slots for slidably receiving lugs formed on edges of the traversing sheet elements perpendicular thereto. 8. The storage rack of claim 7 characterized in that the rack has a bottom plate and a top plate for the lattice arrangement and that each two corresponding narrow sheet elements forming a partition wall are provided close to the lower edge and upper edge with rectangular openings facing each other, for receiving each of which a fitting lugs of a block mounted between the two narrow sheet elements within said interstice, said block being pulled toward the bottom plate or the top plate, respectively, by means of a screw. 9. The storage rack of claim 8, characterized in that at least one of the bottom plate and the top plate is provided with slits for receiving edges of the sheet elements. 10. The storage rack of claim 9, characterized in that within the space between each two lugs of a row belonging to two sheet elements forming a partition wall and extending through the outer wall, a beam is provided, said beam being connected to the lugs, said beams at the lower part and at the upper part being screwed to the bottom plate and the top plate, respectively. 11. A storage rack for fuel elements of nuclear reactors, consisting of a sheet metal lattice arrangement, constituting a plurality of adjoining equal vertical storage compartments of mainly rectangular cross section, each compartment storing one fuel element, characterized in that said lattice arrangement is constructed of a plurality of wide and narrow sheet elements, together forming walls of the compartments, said wide sheet elements being positioned mainly perpendicular to said narrow sheet elements and extending to define walls of a plurality of compartments and said narrow sheet elements extending over the width of one storage compartment, said narrow sheet elements being provided at their opposite longitudinal edges with lugs, said wide sheet elements having slots for receiving the lugs, each lug having a transverse opening for receiving a wedge, said opening after fitting of a lug through a respective one of said slots being positioned with a lug portion at the opposite side of the adjoining wide sheet elements, and a wedge insertable into said opening for tightly pressing each longitudinal edge of the narrow sheet element against the side of the adjoining wide sheet element. 12. The storage rack of claim 11, characterized in that narrow sheet elements extending between two wide sheet elements are offset with respect to narrow sheet elements in each adjoining space between one of said two wide elements and a succeeding wide sheet element. 13. The storage rack of claim 11, characterized in that adjoining storage compartments are separated from each other by walls consisting of two parallel closely spaced narrow and wide sheet elements and in that the lugs and wedges are situated within the space between two parallel closely spaced wide sheet elements. |
description | This application is a division of Ser. No. 10/401,284, filed Mar. 27, 2003, now U.S. Pat. No. 6,888,809, which claims priority from U.S. Patent Application Ser. No. 60/377,784, filed May 3, 2002. This invention relates to reactor pressure vessels (RPVs) in pressurized water nuclear reactors (PWRs). It relates particularly to RPVs in which a small fraction of inlet coolant water is diverted from the main coolant in order to cool the RPV heads. A PWR generally includes a closed loop of pressurized coolant water to transfer heat energy from fuel assemblies in the core region of a RPV to a secondary water system employed to generate steam. The closed loop operates at pressures of up to about 2250 psi or more and at temperatures of up to about 650° F. or more. The coolant water may be heated about 60° F. (e.g., from about 550° F. to about 610° F.) in a RPV and then cooled an equivalent amount by the secondary water system. After decades of operation at high temperature and pressure, the wetted metal surfaces of RPVs (which generally are fabricated of stainless steel and nickel base alloys) contacted by the circulating coolant water are experiencing stress corrosion cracking. One well recognized method of reducing the susceptibility of metals to stress corrosion is to reduce the temperature of the wetted metal surface. Tests have shown that crack initiation times can be reduced significantly by reducing temperatures of the RPV heads by just 10° F. Accordingly, and in addition to other RPV thermal-hydraulic modifications, RPV components have been redesigned or modified to divert relatively cool inlet coolant water away from the main coolant water path and toward the RPV heads in order to cool the RPV heads. Thus, RPVs may have coolant water flow holes machined in the flanges of core barrel assemblies and upper support assemblies to provide a flow path for the diverted inlet coolant water. See, e.g., U.S. Pat. Nos. 5,325,407 and 4,786,461. The inventors have found that, although the flow patterns of U.S. Pat. Nos. 5,325,407 and 4,786,461 will provide the desired cooling effects, the diverted coolant flow may not provide the expected flow of coolant water into the RPV heads in all circumstances. Specifically, it has been found that the flow of coolant water through the holes in the upper support plates can induce coolant water in the RPV heads above the upper support assemblies to leak back into the space between the flanges and dilute the diverted coolant water somewhat like a jet pump aspirating surrounding fluid. In one test, the loss factor (“k”) of Bernoulli's equationPressure Differential=k(Velocity)2/2(gravitational constant)was determined to be 1.6 where a loss factor of 1.1 was employed in the design of the holes. Thus, the quantity of diverted flow of coolant water into the RPV head and its temperature may not be sufficient to cool the head in accordance with the design. It is therefore an object of this invention to provide an improved RPV design for controlling the flow of diverted coolant water into the RPV heads. It is another object of this invention to reduce the susceptibility of coolant water leakage from the RPV head region back into the space between the core barrel support assembly flanges and the upper support assembly flanges. It has been discovered that the foregoing objects can be attained in a reactor pressure vessel (RPV) for containing nuclear fuel assemblies in coolant water by providing: a vessel body having an internal support ledge; a core barrel assembly having a flange supported on the internal support ledge for supporting the fuel assemblies, the core barrel assembly flange having holes extending between a lower flange surface and an upper flange surface, each hole having a cross-sectional area; and an upper support plate having a flange with a lower flange surface and an upper flange surface, the lower surface of the upper support assembly flange disposed above the upper surface of the core barrel assembly flange, the upper support plate flange having holes extending between its lower surface and its upper surface and aligned with the holes in the core barrel assembly, each upper support plate flange hole having a cross-sectional area, wherein the cross-sectional area of the upper support plate flange hole is different from the cross-sectional area of the aligned core barrel assembly flange hole. Advantageously, the relative sizes of the aligned holes in the flanges can be designed in RPVs to better control the inlet coolant water that is diverted from the main stream and to reduce the backflow leakage. In a preferred practice of the present invention, existing RPVs can be backfit to utilize existing holes in the core barrel assembly flanges having cover plates blocking the holes. In this practice, the RPV has: a vessel body having an internal support ledge; a core barrel assembly having a flange supported on the internal support ledge for supporting the fuel assemblies, the core barrel assembly flange having holes extending between a lower flange surface and an upper flange surface with cover plates welded to the bottom surface of the core barrel assembly flange extending under the holes, each cover plate having a hole extending therethrough with a each cover plate hole having a cross-sectional area; and an upper support plate having a flange with a lower flange surface and an upper flange surface, the lower surface of the upper support plate flange disposed above the upper surface of the core barrel assembly flange, the upper support plate flange having holes extending between its lower surface and its upper surface and aligned with the holes in the core barrel assembly, with each upper support plate flange hole having a cross-sectional area, wherein the cross-sectional area of each upper support plate flange hole is different from the cross-sectional area of the hole in the cover plate extending under the aligned core barrel assembly flange hole. In another preferred practice of the present invention, existing RPVs can be backfit with exhaust nozzles to effectively increase the cross-sectional area of the upper support plates. In this practice, the RPV has: a vessel body having an internal support ledge; a core barrel assembly having a flange supported on the internal support ledge for supporting the fuel assemblies, the core barrel assembly flange having holes extending between a lower flange surface and an upper flange surface, each hole having a cross-sectional area; and an upper support plate having a flange with a lower flange surface and an upper flange surface, the lower surface of the upper support plate flange disposed above the upper surface of the core barrel assembly flange, the upper support plate flange having holes extending between its lower surface and its upper surface and aligned with the holes in the core barrel assembly, the upper support plate having exhaust nozzles extending from the holes and above its upper surface, each exhaust nozzle having an exit port having an exit area that is different from the cross-sectional area of the aligned hole in the core barrel assembly. Other and further objects of this invention will become apparent from the following detailed description and the accompanying drawings and claims. FIG. 1 generally illustrates the various components of a reactor pressure vessel (RPV) in a pressurized water reactor (PWR) for which this invention is applicable. The RPV is comprised generally of a cylindrical steel pressure vessel body 1 having a removable top head 2, each provided with cooperating flanges that can be bolted tightly together by a plurality of studs 3 and nuts 4 when the PWR is in service. The removable top head 2 has a plurality of penetration tubes 5 to accommodate control rod drive mechanisms 6 or to function as instrumentation ports for thermocouples or other control instrumentation. The susceptibility of these penetrations 5 to stress corrosion cracking can be reduced by reducing their temperature as much as reasonably possible. One method of reducing their temperature is to reduce the temperature of the coolant water in the vicinity of the penetrations. The vessel body 1 has at least one coolant water inlet nozzle 7, at least one coolant water outlet nozzle 8, and a cylindrical core barrel assembly 9 for supporting fuel assemblies 10 in the core region of the RPV. The coolant water generally enters the vessel body 1 through the inlet nozzle 7 and flows down the annulus between the core barrel assembly 9 and the inner wall of the vessel body 1. The coolant water then turns upwardly at the bottom of the pressure vessel body 1 and flows upwardly through the fuel assemblies 10, a core support plate 12 and out of the vessel body 1 through the outlet nozzle 8 to a steam generator (not shown). The vessel body 1 also has an upper support assembly including an upper support plate 11 for supporting the control rod drive mechanisms 6 and control rod guide tubes 13 for guiding the control rod drive mechanisms 6. The upper support assembly is also designed to permit coolant water to circulate between RPV head region and the region above the fuel assemblies 10 in order to cool the RPV head 2 and its penetrations. In a desirable flow pattern for cooling the RPV head 2, the diverted coolant water flows from the RPV inlet nozzle 7, through holes in the flanges of the core barrel assembly 9 (shown in FIG. 24) and the upper support plate 11, into the RPV head 2, downwardly through flow spaces in the upper support assembly, and then into the region above the fuel assemblies 10. FIG. 2 generally shows a first preferred embodiment of the present invention. This figure generally shows the RPV region where the core barrel support assembly 9 and the upper support plate 11 are supported by the vessel body 1. The vessel body 1 has an internal support ledge 20 that supports a core barrel assembly flange 22. The flange 22 has holes (represented by hole 24) extending from a lower surface 26 to an upper surface 28. The upper support plate 11 has a flange 32. The upper support plate flange 32 has holes (represented by hole 34) extending from a lower surface 36 to an upper surface 38. Importantly, in the present invention, the holes in the flanges 22 and 32 are aligned such that the hole 24 in the core barrel assembly flange 22 is directly under the hole 34 in the upper support assembly flange 32. A spring 40 may be disposed between the core barrel assembly flange 22 and the upper support assembly flange 32 for supporting the upper support assembly 11 on the core barrel assembly 9. As is shown in FIG. 2, the cross-sectional areas of the aligned holes in the flanges 22 and 32 of FIG. 2 (i.e., the areas taken parallel to the upper and lower surfaces) are different from each other, which enables the RPV designer to selectively design the pressure drops and velocities of the coolant water flowing through the holes 24 and 34 from the annular RPV region 42 communicating with the coolant inlet nozzle 7 into to the RPV region 44 in the RPV head 2. Preferably, the holes 34 are larger in diameter (and therefore have a larger cross-section) than aligned holes 24 for reducing the relative velocity of the coolant water and therefore the entrainment of coolant water leaking around the periphery 46 of the upper support plate flange 32. As is also shown in FIG. 2, the holes 24 and 34 have substantially vertical sidewalls so that the cross-sectional areas within the holes are constant. In one application of this embodiment, the diameters of holes 24 are approximately 2¼ inches and the diameters of holes 34 are approximately 3¼ inches. In other embodiments of the present invention, the sidewalls of either holes 24 and/or holes 34 may be tapered. Thus, the holes 24 and 34 may diverge such that the cross-sectional hole areas at the lower surfaces 26 and 36 are smaller than the cross-sectional hole areas at the upper surfaces 28 and 38. Advantageously, this design will tend to reduce the expansion pressure drop of the coolant water as it flows from the flanges 22 and 32. In other embodiments of the present invention, the holes 24 and 34 may be oblong slots instead of circular. FIG. 3 illustrates a second embodiment of the present invention in a backfitted RPV having internally threaded lifting holes 52 for receiving the threaded ends of lift rods (not shown), which are employed to transfer the core barrel assembly 9 into the RPV. Cover plates 54 blocking the lifting holes 52 are welded to the bottom surface 26 of the core shroud assembly flange 22 at the time of manufacture for the specific purpose of preventing coolant water flow. The lifting holes 52 are generally aligned with holes 56 in the upper support plate flange 32, which holes 52 and 56 may have the same or different cross-sections. In this embodiment of the present invention, holes 58 are drilled in the cover plates 54, which holes 58 have a different cross-sectional area than the cross-sectional area of the aligned holes 56 in the upper support assembly flange 32. As shown in FIG. 3, the holes 56 in the upper support plate flanges 32 are larger than the holes 58 in the cover plates 54. Advantageously, the flow of coolant water may be controlled by the relative cross-sectional areas of these holes in combination with the threaded configuration of the sidewalls of the core barrel assembly flange lifting holes 52. FIG. 4 illustrates a third embodiment of the present invention in a backfitted RPV having a core barrel assembly flange 22 and an upper support plate flange 32 with aligned holes 62 and 64, respectively. The holes 62 and 64 may have the same or different cross-sections. As shown, exhaust nozzles 66 (or diverging nozzles) are inserted in the holes 64 in the upper support plate flanges 32. Preferably, each exhaust nozzle 66 has one or more exit ports 68 with a total flow area that is different from the cross-sectional area of the aligned hole 62 in the core barrel assembly flange 22. Preferably, the flow area of the exit port 68 is sufficient to reduce the coolant water flow to about 25% of the flow through the hole 62 in the core barrel assembly flange 22. In addition to controlling velocities and pressure drops, exhaust nozzles 66 discharge the coolant water above the bottom surface of the RPV head 2. While present preferred embodiments of the present invention has been shown and described, it is to be understood that the invention may be otherwise variously embodied within the scope of the following claims of invention. |
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summary | ||
047612614 | abstract | A novel liquid nuclear reactor is described which comprises a reactor vessel that is connected through upper and lower liquid conduit means to one or more satellite tanks that contain a heat exchanger means and pump means. |
claims | 1. An apparatus for measuring an aerial image, the apparatus comprising:a movable unit adapted to move a reflective extreme ultra-violet (EUV) mask disposed thereon in an x-axis and/or y-axis direction;an X-ray mirror arranged on the movable unit, the X-ray mirror being adapted to selectively reflect a coherent EUV light having a selected wavelength;a zoneplate lens that is located between the movable unit and the X-ray mirror, the zoneplate lens being adapted to focus the coherent EUV light on a portion of the reflective EUV mask; anda detector arranged on the movable unit, the detector being adapted to sense energy of the reflected coherent EUV light when the focused coherent EUV light is reflected by the portion of the reflective EUV mask,wherein numerical apertures of the zoneplate lens and detector are based on parameters of a scanner used to form a pattern through the EUV mask, the parameters of the scanner including a numerical aperture, an off-axis degree, and a reduction magnification. 2. The apparatus as claimed in claim 1, further comprising an aperture between the reflective EUV mask and the detector. 3. The apparatus as claimed in claim 1, wherein the X-ray mirror comprises a multi-layer structure including at least one molybdenum layer and at least one silicon layer, which are alternately arranged. 4. The apparatus as claimed in claim 1, further comprising a EUV light generator, the EUV light generator comprising:a high power femtosecond laser adapted to output a high power femtosecond laser beam;a gas cell adapted to generate the coherent EUV light having a selected wavelength from the high power femtosecond laser; anda lens adapted to focus the high power femtosecond laser beam on the gas cell. 5. The apparatus as claimed in claim 4, wherein the gas cell is filled with a neon gas so as to optimize a production efficiency of a coherent EUV light having a wavelength of 13.5 nm. 6. The apparatus as claimed in claim 4, wherein the X-ray mirror is adapted to reflect the coherent EUV light emitted from the EUV light generator toward the portion of the reflective EUV mask at an angle of about 4° to about 8° with respect to a normal line of the reflective EUV mask. 7. The apparatus as claimed in claim 1, wherein the zoneplate lens is adapted to focus the reflected coherent EUV light on the portion of the reflective EUV mask at an angle of about 4° to about 8° with respect to a normal line of the reflective EUV mask. 8. The apparatus as claimed in claim 1, further comprising a computing unit adapted to reconstruct an image of the reflective EUV mask based on energy sensed by the detector. 9. The apparatus as claimed in claim 1, wherein numerical apertures of the zoneplate lens and detector are based on incidence and reflection angles of the coherent EUV light on and from the reflective EUV mask. 10. An apparatus for measuring an aerial image of a pattern corresponding to a semiconductor pattern to be formed by scanning the pattern using a scanner, the apparatus comprising:a zoneplate lens arranged on a first side of an extreme ultra-violet (EUV) mask including the pattern, the zoneplate lens adapted to focus EUV light on a portion of the EUV mask at a same angle as an angle at which the scanner will be disposed with respect to a normal line of the EUV mask; anda detector arranged on a second side of the EUV mask and adapted to sense energy of the EUV light from the EUV mask,wherein numerical apertures of the zoneplate lens and detector are based on parameters of a scanner used to form a pattern through the EUV mask, the parameters of the scanner including a numerical aperture, an off-axis degree, and a reduction magnification. 11. The apparatus as claimed in claim 10, further comprising a movable unit on which the EUV mask is arranged, the movable unit being adapted to move the EUV mask in an x-axis direction and/or an y-axis direction. 12. The apparatus as claimed in claim 10, wherein the EUV mask is a reflective EUV mask including a reflective material. 13. The apparatus as claimed in claim 12, wherein the detector is adapted to sense energy of reflected EUV light that is reflected from the reflective EUV mask. 14. The apparatus as claimed in claim 10, further comprising an EUV light generator and an X-ray mirror adapted to selectively reflect the EUV light from the EUV light generator. 15. The apparatus as claimed in claim 14, wherein the EUV light generator includes a high power femtosecond laser. 16. The apparatus as claimed in claim 10, wherein the EUV mask is a transmissive EUV mask. 17. The apparatus as claimed in claim 16, wherein the detector is adapted to sense energy of transmitted coherent EUV light that is transmitted through the transmissive EUV mask. 18. An apparatus for measuring an aerial image, the apparatus comprising:a light source configured to emit extreme ultra-violet (EUV) light;a movable unit adapted to move a reflective EUV mask disposed thereon in an x-axis and/or y-axis direction;a zoneplate lens configured to focus the EUV light on the EUV mask at a same angle as an angle at which a scanner will be disposed with respect to a normal to the EUV mask;a detector arranged on the movable unit, the detector being adapted to sense energy of RUV light reflected by the portion of the reflective EUV mask, wherein a ratio of a numerical aperture of the detector to a numerical aperture of the zoneplate lens equals an off-axis degree of the scanner used to form a pattern through the EUV mask. 19. An apparatus for measuring an aerial image, the apparatus comprising:a movable unit adapted to move an extreme ultra-violet (EUV) mask disposed thereon;a mirror arranged on the movable unit, the mirror being adapted to reflect an EUV light;a zoneplate lens located between the movable unit and the mirror, the zoneplate lens being adapted to focus the EUV light on a first region and on a second region of a first side of the EUV mask; anda detector arranged on the movable unit, the detector being adapted to sense energy of the EUV light reflected from the first region and second region of a second side of the EUV mask. 20. The apparatus as claimed in claim 19, wherein a numerical aperture of the zoneplate lens is based on parameters of a scanner used to form a pattern through the EUV mask, the parameters of the scanner including a numerical aperture and a reduction magnification. 21. The apparatus as claimed in claim 20, wherein NAdetector=NAscanner/n, where NAdetector denotes a NA of the detector, NAscanner denotes a NA of the scanner, and n denotes a reduction magnification of the scanner. 22. The apparatus as claimed in claim 19, wherein a ratio of a numerical aperture of the detector to a numerical aperture of the zoneplate lens equals an off-axis degree of the scanner used to form a pattern through the EUV mask. 23. An apparatus for measuring an aerial image of a pattern corresponding to a semiconductor pattern to be formed by scanning the pattern using a scanner, the apparatus comprising:a zoneplate lens arranged on a first side of an extreme ultra-violet (EUV) mask including the pattern, the zoneplate lens adapted to focus an EUV light on different regions of the EUV mask; anda detector arranged on the first side of the EUV mask and adapted to sense energy of the EUV light reflected separately from the different regions of the EUV mask, the EUV light reflected from the different regions of the EUV mask passing outside the zoneplate lens. 24. The apparatus as claimed in claim 23, wherein numerical aperture of the zoneplate lens is based on parameters of a scanner used to form a pattern through the EUV mask, the parameters of the scanner including a numerical aperture and a reduction magnification. 25. The apparatus as claimed in claim 24, wherein NAdetector=NAscanner/n, where NAdetector denotes a NA of the detector, NAscanner denotes a NA of the scanner, and n denotes a reduction magnification of the scanner. 26. The apparatus as claimed in claim 23, wherein the zoneplate lens is adapted to focus the EUV light on a portion of the EUV mask at a same oblique angle as an oblique angle at which the scanner will be disposed with respect to a normal line of the EUV mask. |
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040381340 | abstract | An improved pressurized water-cooled reactor system which includes a reactor core pressure vessel; coolant lines coupled to the pressure vessel; vertically arranged steam generators having the housings thereof coupled to the coolant lines; and coolant pumps associated with the steam generators and coupled to the coolant lines. Each of the pumps has a vertically arranged impeller portion driven by an electric motor for circulating coolant between the pressure vessel and tube sheet type heat exchangers in the steam generators, at least the impeller portion of each pump being disposed within a chamber associated with each steam generator which is communicative with the coolant lines. The vessel, the coolant lines and the steam generators including the impeller chamber are each individually encapsulated by rupture-proof safety housings (encasements) forming a protection system for the primary loop components. |
046997539 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The invention will be described in connection with the detachable control console of the nuclear reactor refueling machine described in U.S. Pat. No. 4,511,531 which is hereby incorporated by reference into this application to provide a complete description of such a machine and its operation, but which is not required for a full understanding of the present invention. A schematic representation of such a refueling machine 1 is shown in FIG. 1. The machine 1 includes a bridge 2, a trolley 4 and a hoist 6 supported on a mast 8. It straddles a reactor 10 which is located under 20 to 30 feet of water 12 at the bottom of a pit 14 in a containment defined by massive walls 16. The bridge 2 is mounted on wheels 18 which ride on rails 20 extending along the sides of the pit 14 on the walls 16. The wheels 18 are driven by a bridge motor 22 through a gear reducer 24 and drive shafts 26. Similarly, the trolley 4 is mounted on wheels 28 for movement on rails 30 extending longitudinally along the bridge 2. A trolley motor 32 drives the wheels 28 through a gear reducer 34 and drive shaft 36. The hoist 6 is driven by hoist motor 38 to raise and lower an elongated tube and an inner mast (neither shown) inside a guide mast 40. The details of the elongated tube and the inner mast which are positioned by the hoist 6 are described in U.S. Pat. No. 4,511,531. The bridge, trolley, and hoist motors, 22, 32 and 38 are equipped with pulse generators 42, 44 and 46 respectively which generate pulse feedback signals representative of incremental rotation of the motor, and hence of movement of the associated refueling machine component The control console 48, which is mounted on the trolley 4 during refueling operations, responds to operator inputs and feedback from the refueling machine, such as the feedback signals from the pulse generators 42, 44, and 46, to generate various control signals for the operation of the refueling machine. A CRT 50 on the console provides the operator with a visual display of pertinent refueling machine parameters. The control console 48 is removed from the refueling machine during operation of the reactor at power to protect the solid state electronic equipment in the console from the high levels of radiation, temperature, and humdity present in containment. As shown in FIG. 2, a refueling machine simulator 52 is connected to the detached control console 48 by a group of cables 54, 56, and 58 having quick disconnects 60 at each end. These cables replace the cables by which the control console is connected to the refueling machine 1 during refueling operations. Power cords 62 and 64 provide 220 volt three phase and 115 VAC power respectively to the simulator 52, which it will be seen, supplies these services to the control console through cable 54. The simulator 52 includes a steel enclosure 66 which houses a single three phase simulator motor 68 which drives a single rotary pulse generator 90. The simulator 52 also includes a number of relays 72, the functions of which will be discussed in detail below, and associated circuitry (not shown in FIG. 2). Switches 74 on the front of the simulator 52 control various test modes of the simulator. A removable lid 76 on the enclosure 66 is provided with a clear LEXAN window 78 for ease in viewing the moving parts. The simulator 52 is a compact self-contained unit which fits easily in a small suitcase for transportation and storage. As shown in the schematic circuit diagram of FIG. 3, three phase power applied to the simulator 52 through power cord 62 is supplied to the control console 48 by four leads of cable 54. The 120 VAC power received through power cord 64 is used to power a number of relays 72 in the simulator 52 and is also supplied to the control console through leads in cable 54. The 120 VAC is used by the control consoles to power its controls and logic circuits while the 220 volt three phase power is used to power a variable frequency, pulse width modulated motor controller. Logic circuits within the control console respond to operator inputs to control the motor controller and to connect the three phase drive signals D.sub.B, D.sub.T or D.sub.H generated by the motor controller to one of three sets of output leads 80, 82, 84 in cable 54 for powering the bridge, trolley or hoist motors 22, 32 or 38 respectively on the refueling machine 1. However, all of these drive signals are connected through a network 86 within the simulator to the single simulator motor 68. The logic circuits of the control console 48 also generate brake or control signals B.sub.B, B.sub.T or B.sub.H for the bridge, trolley or hoist motor respectively on the refueling machine, as appropriate, simultaneously with the selected drive signals. These signals are connected through separate leads 88, 90, and 92 in cable 54 to the coil of a bridge brake relay 72B, a trolley brake relay 72T, and a hoist brake relay 72H respectively in the simulator 52 to energize the same. Each of these brake or control signals is also connected through a lead 94 to the coil of an electrically releasable brake 96 on the signal simulator motor 68. Thus, generation of any of the brake signals releases the brake on the single simulator motor, but energizes only the designated relay coil. Each of the relays 72B, H and T has four sets of make contacts, 72B-1 to 4, 72T-1 to 4 and 72H-1 to 4, with one contact from each relay connected in parallel in one of four leads, 70A, B, D and E carrying the output signals generated by the single rotary pulse generator 70. The pulse generator 70 generates pulse signals on the leads 70A, B, D and E at a rate and in a pattern determined by the direction and rate of rotation of the single simulator motor 68. These respective signals are transmitted back to the control console 48 as feedback signals. When the contacts 72B-1 to 4 of the bridge relay 72B are closed, the pulse signals are routed back to the control console 48 as feedback signals FB through console input leads 98 in cable 56. These are the same leads over which the control console 48 receives pulse signals from the pulse generator 42 associated with the bridge motor 22 when the console is connected to the refueling machine 1. Similarly, with the contacts 72T-1 to 4 closed the pulse signals are directed to the console as trolley motor feedback signals FT over leads 100 in cable 58, and with contacts 72H-1 to 4 closed, as hoist motor feedback signals FH over leads 102 in cable 58. As mentioned, the refueling machine of U.S. Pat. No. 4,511,531 includes an inner mast which has grippers on the lower end for grasping fuel assemblies so that they may be lifted out of the reactor core and transported about in containment. The grippers are actuated by pneumatic cylinders which are controlled by solenoid valves. When the gripper is to engage a fuel assembly, the control console sends a fuel assembly gripper engage signal FE over an output lead in cable 54 to a fuel assembly gripper engage solenoid. When the gripper moves to the engage position, a limit switch is closed to generate a feedback signal EF which is sent back to the console on an input lead in cable 56. When the fuel assembly is to be disengaged, a fuel assembly gripper disengage signal FD is sent to a disengage solenoid which results in generation of a fuel assembly gripper disengage signal DF when the corresponding limit switch is closed. The machine of U.S. Pat. No. 4,511,531 includes a second set of grippers which is mounted on a tube which telescopes into the mast. This set of grippers, which is also pneumatically actuated, is designed for gripping other reactor components such as control rod clusters. The pneumatic cylinder is actuated to the engage position by a solenoid in response to a control road cluster engage signal, CE, which generates an engage signal EC when the corresponding limit switch is closed. Likewise, a control rod cluster disengage signal CD energizes the disengage solenoid which produces a disengage signal DC as the limit switch closes. The tube carrying the control rod cluster grippers in U.S. Pat. No. 4,511,531 is raised and lowered by the hoist 6. A pivotable stop plate on the inner mast carrying the fuel assembly grippers is actuated to a position wherein lugs on the tube engage the stop plate to lift the inner mast when a plate engage solenoid is energized by a plate engage signal PE from the control console to actuate a pneumatic cylinder. The stop plate is pivoted to a position where the plate is clear of the lug so that the inner mast is not lifted with the inner tube when a disengage solenoid is energized by a disengage signal PD. Limit switches generate corresponding engage and disengage signals EP and DP when the stop plate reaches the respective positions. When the simulator 52 is connected to the control console, the FE and FD signals carried by leads in cable 54, energize the coils of fuel assembly gripper engage and disengage relays 72FE and 72FD respectively. Similarly, the signals CE and CD, associated with the control rod grippers, and PE and PD, associated with the stop plate, energize the coils 72CE, 72CD, 72PE and 72PD in the simulator 52 respectively. A set of break contracts associate with each of these relays, such as 72FE-1, completes a circuit between leads in the cables 56 and 58 connected to the console to generate the respective feedback signals set forth above. The refueling machine is provided with a pair of load-cells which have an electrical resistance which is proportional to the load on the hoist. The signals generated by these load cells provide an indication of whether the weight of the fuel assembly or control rod cluster is supported by the hoist and whether there is any obstruction to free movement of the supported component. These load cells are simulated by potentiometers 104 and 106 which are connected to the control console through leads in cable 56. The potentiometers can be set to simulate a load condition for testing this feature of the control console. Another potentiometer 108 is provided to simulate the pneumatic system pressure transducer and thereby test the console's reading of the transducer output. Several switches provide means for checking certain functions of the control console 1. Switch 110 tests the console's response to a geared limit switch which protects against overtravel of the hoist. Switch 112 simulates a hoist motor overheat condition. Switch 114 indicates whether the mast has rotated out of its normal position. Since the positions of the bridge, hoist and trolley are determined by counting pulses indicative of incremental movement, check switches 116, 118 and 120 are mounted at a known location in the path of each of these components to provide a means for checking the calculated positions and to reset them if they are out of synchronization. The refueling machine can transport fuel assemblies to and from a transfer area where they can be removed from or introduced into containment by a transfer system. An interlock system prevents interference between the refueling machine and the transfer system. A lamp 122 on the simulator checks the generation of the interlock signal by the control console 48. In operation, the simulator is connected to the control console 48 by cables 54, 56 and 58 and power cords 62 and 64 are connected to a power source. When used as a simulator, the operator operates the controls of the console to produce the desired simulated refueling machine movement. For instance, if movement of the bridge is to be simulated, the proper console controls are actuated to generate a bridge brake signal, B.sub.B on output lead 88 in cable 54. This energizes the simulator motor brake coil 96 through lead 94 to release the brake. It also energizes the coil of relay 72B to close the contacts 72B-1 to 4. The control console also connects the motor controller to output leads 80 in cable 54 associated with the bridge motor and controls the phase and width of pulses generated by the controller to produce a three phase signal, D.sub.B, which is connected through network 86 to the single simulator motor 68. As the motor begins to turn in the direction and at the rate dictated by the signal D.sub.B, the pulse generator 70 generates a pattern of pulses on the leads 70A, B, D and E porportional to this movement. This pulse signal is fed back to the console 48 as feedback signal FB on the console input leads 98 in cable 56. The console 48 tracks movement of the bridge by counting the pulses in the signal FB and displays bridge position on the CRT display 50. When movement of the trolley is commanded, the console generates the signal B.sub.T on output lead 90 in cable 54 to release the simulator motor brake by energizing coil 96. The signal B.sub.T also energizes coil 72T to direct pulses generated by pulse generator 70 through contacts 72T-1 to 4 to console input leads 100 as feedback signals FT. Movement of the trolley simulated by the motor 68 and pulse generator 70, and represented by the feedback signal FT, is followed by the console and also presented on the CRT display 50. When hoist movement is desired, the operator actuates the console controls to generate a hoist brake signal B.sub.H which is applied to the simulator through lead 92 of cable 54 to energize the coil 96 of the simulator motor brake and the coil of relay 72H. Make contacts 72H-1 to 4 of relay 72H direct the pulses from pulse generator 70 to input leads 102 in cable 58 for transmission to the console as the hoist motor feedback signal FH. Movement of the hoist is also tracked by the console and displayed on the CRT 50. Thus, it can be seen that a single simulator motor 68 and pulse generator 70 simulate all three refueling machine motors. The separate leads which direct the three phase drive signals to the three separate refueling machine motors are all connected to operate the one simulator motor. In addition, the brake signals for each of the refueling machine motors which are each carried by a separate lead are also connected to energize the one brake coil for the single simulator motor. Whereas, in the refueling machine, these signals only release the appropriate brake; in the simulator they each energize a separate relay which closes contacts to direct the pulses generated by the single pulse generator to the appropriate separate console input line associated with the indicated refueling machine motor. When the operator provides an input to the console 48 for the grippers to engage a fuel assembly or a control rod cluster, a signal generated by the console energizes the appropriate relay 72FE or 72CE which opens its normally closed contacts 72FE-1 or 72CE-1 to simulate the engagement. When the gripper is to release the fuel assembly or control rod cluster, the console generates a disengage signal which energizes relay 72FD or 72CD to open its normally closed contacts and provide an indication of the disengagement. When a fuel assembly is to be lifted, the console generates a stop plate engage signal, PE, which energizes relay 72PE to generate an indication by opening contacts 72PE-1 of the stop plate being pivoted into position to couple the inner mast carrying the fuel assembly grippers to the tube lifted by the hoist. When the console generates a disengage stop plate signal, PD, the contacts of relay 72PD open to simulate pivoting of the stop plate to a position where the inner mast is disengaged from the tube connected to the hoist. The remaining features of the simulator are used during testing of the console. The load cell potentiometers and the pressure transducer potentiometer can be used to test the console load and pressure measurements. The various switches can be actuated to verify the response by the control console to the test conditions. The simulator of this invention may be used to troubleshoot the console when it is installed on the refueling machine as well as when the console has been removed from containment. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof. |
053405056 | abstract | A method for dissolving radioactively contaminated surfaces of metal articles using a reagent of HBF.sub.4 acid with the addition of at least one oxidation agent, preferably hydrogen peroxide H.sub.2 O.sub.2, for the efficient decontamination of radioactively contaminated metal articles. An optimum mixture of this reagent was about 5% HBF.sub.4 acid with the addition of about 0.5% by volume of H.sub.2 O.sub.2. Radioactively contaminated lead plates, for example, were treated by this reagent, and the contaminated solution was used as an electrolyte without any further additive. The contaminated lead or lead oxide is deposited at the anode or cathode decontaminating the solution which may be returned to the process. If, instead of lead, the metal is copper, nickel, steel, silver or mercury or their alloys, the method for dissolving radioactively contaminated surfaces can be executed in the same way with the same reagent. |
abstract | A monochromator (1) for a charged particle optics, in particular, for electron microscopy, comprises at least one first deflection element (2, 3) with an electrostatic deflecting field (2′, 3′) for generating a dispersion (4) in the plane (5) of a selection aperture (6) to select charged particles of a desired energy interval (7) and at least one second deflection element (8, 9) with an electrostatic deflecting field (8′, 9′) which eliminates the dispersion (4) of the at least one first deflecting field (2′, 3′). A radiation source (17) comprises such a monochromator (1). High monchromatism without intensity contrasts caused by defects of the slit aperture is thereby achieved in that the deflection elements (2, 3, 8, 9) have a design other than spherically shaped and their electrodes (24, 25) are given a potential (φ+, φ−) such that the charged particles (xα, yβ) which virtually enter the image of the radiation source (17) at different respective angles (α, β) in different sections (x, y), are differently focused such that charged particles (xα, yβ) of one energy are point focused (10, 10′, 10″) exclusively in the plane (5) of the selection aperture (6), since zero-crossings (11, 12) of the deflections (A) of the charged particles (xα, yβ) of the different sections (x, y) only coincide there at the same axial position (z, E). |
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048470368 | claims | 1. In a water-filled pool having a nuclear reactor fuel assembly therein, the nuclear reactor fuel assembly having fuel rods with spaces therebetween, an apparatus for the inspection of the nuclear reactor fuel assembly, the apparatus comprising a dive housing having walls, a rotary transmission extending through one of said walls of said housing toward the surface of water in the pool, two vertically spaced apart lever linkages jutting out from said rotary transmission, each of said lever linkages having an end facing away from said rotary transmission, carrying plates each being connected to a respective one of said ends of said lever linkages, a tie bar interconnecting said carrying plates, and probes being disposed at a plurality of vertical levels along said tie bar for simultaneously moving laterally into a space between two of the fuel rods. 2. Apparatus according to claim 1, wherein said rotary transmission has an end facing away from said housing, and including a cover plate in which said end of said rotary transmission is guided, and struts connected between said cover plate and said housing. |
description | 1. Field of the Invention The present invention relates to a fuel rod testing apparatus for a nuclear fuel assembly and, more particularly, to a fuel rod testing apparatus for effectively performing a helium leakage test and macrography on a fuel rod that is finally assembled as a unit part in the process of manufacturing the fuel rod. 2. Description of the Related Art Nuclear reactors are facilities for artificially controlling a fission chain reaction of a fissionable material in order to use thermal energy generated from nuclear fission as power. Referring to FIG. 1, a nuclear fuel assembly includes spacer grids 2 into which fuel rods are inserted, numerous guide thimbles 3 fixed to the spacer grids 2, an upper end fitting 3 fastened to upper ends of the guide thimbles 2, and a lower end fitting 4 fastened to lower ends of the guide thimbles 2. Each fuel rod is supported by dimples and springs formed on each spacer grid 2. The fuel rod is made up of a cladding tube, end plugs, pellets, and springs. The fuel rod is manufactured by enriching uranium-235 of 2 to 5% to form a cylindrical nuclear fuel pellet of about 5 g, charging the pellets into the cladding tube, inserting the springs, and sealing the cladding tube using the end plugs. In the process of charging and sealing the pellets into the cladding tube, the cladding tube is filled with an inert gas such as pressurized helium, thereby reducing and preventing oxidation of the pellets. For example, a nuclear fuel rod having a structure capable of being effectively filled with helium is proposed in Korean Unexamined Patent Application Publication Nos. 10-2006-0134959 (published on Dec. 28, 2006) and 10-1988-0004492 (published on Jun. 4, 1988). In this way, helium introduced into the fuel rod under pressure can reduce or prevent the oxidation of the pellets. Especially, the helium introduced under pressure in the fuel rod for a pressurized water reactor (PWR) functions to relieve compressive stress and creep of the cladding tube, which are generated due to pressure of external cooling water. Thus, a process of testing for leakage of the helium after the fuel rod has been manufactured is required. In regard to this, a method and apparatus for testing for leakage of helium from a nuclear fuel rod are proposed in Korean Unexamined Patent Application Publication Nos. 10-1990-0012289 (published on Aug. 3, 1990). Referring to FIG. 2, the apparatus for testing for leakage of helium from a nuclear fuel rod is provided with a testing chamber 16 having a sealed structure in which it is tested whether or not the helium leaks from the nuclear fuel rod, a first conveyer 36 disposed at one end of the testing chamber 16 to convey the fuel rod to be tested, a second conveyer 38 disposed inside the testing chamber 16, and a third conveyer 40 disposed at the other end of the testing chamber 16 to convey the tested fuel rod. In the helium leakage testing apparatus, the fuel rod is conveyed by the first conveyer 36. Another fuel rod waits for testing outside the testing chamber 16. The fuel rods tested in the testing chamber 16 are discharged to the outside of the testing chamber 16 by the second conveyer 38 installed inside the testing chamber 16 and the third conveyer 40. The discharged fuel rods are transferred to the next process by a transfer line. The fuel rods have a length of about four meters and are conveyed in a horizontal direction. For these reasons, to transfer the fuel rods to the next process in the testing chamber using the conveyer, a long transfer line is required, and thus a wide space for facilities is required. Consequently, efficiency of the process is reduced. Accordingly, the present invention has been made keeping in mind the above problems occurring in the related art, and an object of the present invention is to provide a fuel rod testing apparatus for a nuclear fuel assembly, capable of effectively performing a helium leakage test and macrography on a fuel rod that is finally assembled as a unit part in the process of manufacturing the fuel rod and making efficient use of a space for facilities. To achieve the aforementioned object, there is provided a fuel rod testing apparatus for a nuclear fuel assembly, which includes: a helium leakage testing chamber having a gate at one side thereof so that a fuel rod is horizontally loaded/unloaded in a lengthwise direction and testing whether or not helium leaks from the fuel rod; a fuel rod upward/downward transfer unit that has first and second transfer sections located in front of the gate and horizontally installed on upper and lower stages apart from each other in order to guide the fuel rod loaded into or unloaded from the helium leakage testing chamber, and that vertically drives the first and second transfer sections; and a main frame that has an upper transfer section disposed in parallel in a lengthwise direction of the fuel rod upward/downward transfer unit and having an inclined face toward the fuel rod upward/downward transfer unit, and a lower transfer section installed at a lower portion of the upper transfer section and having an inclined face in an opposite direction of the inclined face of the upper transfer section. Here, the first and second transfer sections may each include at least one drive roller and a plurality of idle rollers. Further, the fuel rod testing apparatus may further include a testing table installed adjacent to the lower transfer section, wherein the fuel rod undergoing a helium leakage test is transferred to the testing table via the lower transfer section. Also, the upper transfer section may further include a stopper member that is allowed to protrude on a transfer path of the fuel rod. In addition, the stopper member may include first and second stoppers that protrude upward from opposite ends of a lever pivotally installed on the upper transfer section, and a rotary driver rotating the lever. The first and second stoppers may be spaced apart a predetermined distance from each other, and protrude from a transfer face of the upper transfer section at different points of time. According to the fuel rod testing apparatus of the present invention, the helium leakage testing chamber is used to test whether or not the helium leaks from the fuel rod, and the fuel rod upward/downward transfer unit has the first transfer section and the second transfer section located below the first transfer section in order to guide the fuel rod loaded into or unloaded from the helium leakage testing chamber, and vertically drives the first and second transfer sections. The main frame has the upper transfer section disposed in parallel in the lengthwise direction of the fuel rod upward/downward transfer unit and having the inclined face toward the fuel rod upward/downward transfer unit, and the lower transfer section installed below the upper transfer section and having the inclined face in the opposite direction of the inclined face of the upper transfer section. For the purpose of a helium leakage test and macrography for long fuel rods, the transfer path is provided in a three-dimensional space rather than a two-dimensional space. Thereby, the helium leakage test and macrography can be effectively performed, and a space for facilities required for processes is minimized so that spatial application can be maximized. Reference will now be made in detail to an embodiment of the present invention with reference to the accompanying drawings. Referring to FIG. 3, a fuel rod testing apparatus for a nuclear fuel assembly according to an embodiment of the present invention includes a helium leakage testing chamber 100 for testing whether or not helium leaks from a fuel rod, a fuel rod upward/downward transfer unit 200 that has a first transfer section 210 and a second transfer section 220 located at a lower portion of the first transfer section 210 in order to guide the fuel rod loaded into or unloaded from the helium leakage testing chamber 100 and that vertically drives the first and second transfer sections, and a main frame 300 that has a upper transfer section 310 disposed in parallel in a lengthwise direction of the fuel rod upward/downward transfer unit 200 and having an inclined face toward the fuel rod upward/downward transfer unit 200 and a lower transfer section 320 installed at a lower portion of the upper transfer section 310 and having an inclined face in an opposite direction of the inclined face of the upper transfer section 310. The helium leakage testing chamber 100 is used to test whether or not the helium leaks from the fuel rod. To this end, a known helium leakage testing apparatus described in the related art may be used. The helium leakage testing chamber 100 is provided with a gate at one side thereof so that the fuel rod can be horizontally transported into or out of the chamber in a lengthwise direction. Further, to allow the fuel rod to be transported into or out of the chamber, the helium leakage testing chamber 100 may be provided therein with a conveyer 110 that can transfer the fuel rod in cooperation with the fuel rod upward/downward transfer unit 200. The fuel rod upward/downward transfer unit 200 is located in front of the gate of the helium leakage testing chamber 100, and includes the first and second transfer sections 210 and 220 that are horizontally installed apart from each other at upper and lower ends thereof in order to guide the fuel rod loaded into or unloaded from the helium leakage testing chamber 100 and that can be vertically driven. In detail, the first transfer section 210 may be made up of a plurality of rollers 211 and 212, preferably at least one drive roller 211 and a plurality of idle rollers 212. The drive roller 211 may include a power moller in which a small motor and a speed reducer are mounted and driven. In the first transfer section 210, the drive roller 211 is driven to load the fuel rod, which is placed on the drive roller 211, into the helium leakage testing chamber 100. Like the first transfer section 210, the second transfer section 220 is also made up of at least one drive roller and a plurality of idle rollers. The fuel rod loaded into the helium leakage testing chamber 100 can be unloaded by the second transfer section 220. Referring to FIG. 4, the first transfer section 210 and the second transfer section 220 are disposed apart from each other in a vertical direction, and are supported on a guide bracket 201. The guide bracket 201 may be vertically driven by a drive unit 230 located at a lower portion thereof. The drive unit 230 may include a pneumatic or hydraulic cylinder or a driving motor. The main frame 300 is a hexahedral beam structure 330 assembled with a plurality of beams, and has the upper transfer section 310 disposed in parallel in a lengthwise direction of the fuel rod upward/downward transfer unit 200 and having a predetermined inclination and the lower transfer section 320 installed at the lower portion of the upper transfer section 310 and having an inclination opposite to that of the upper transfer section 310. The upper transfer section 310 has the inclined face toward the fuel rod upward/downward transfer unit 200. Thus, the fuel rod 1 located at the upper transfer section 310 is rolled toward the fuel rod upward/downward transfer unit 200 along the inclined face of the upper transfer section 310. Referring to FIGS. 4 and 5, in the present embodiment, the upper transfer section 310 may be additionally provided with a stopper member that is allowed to protrude so as to stop the fuel rod 1 on a transfer path. The stopper member may be made up of first and second stoppers 312 and 313 that protrude upward from opposite ends of a lever 311 fastened to the upper transfer section 310 by a hinge h, and a rotary driver for rotating the lever 311. According to an angle of rotation of the lever 311, a point of time at which the first stopper 312 protrudes is opposed to that at which the second stopper 313 protrudes. For example, when the first stopper 312 protrudes from the upper transfer section 310, the second stopper 313 is located below a fuel rod transfer face of the upper transfer section 310, so that the fuel rod moves downward along the upper transfer section 310 without interfering with the second stopper 313. In contrast, when the second stopper 313 protrudes, the first stopper 312 is located below the fuel rod transfer face of the upper transfer section 310, so that the fuel rod moves along the upper transfer section 310 without interfering with the first stopper 312, and can be placed on the first transfer section 210. The first stopper 312 and the second stopper 313 are spaced apart a predetermined distance D almost corresponding to a length of the lever 311. The distance D between the first stopper 312 and the second stopper 313 corresponds to the number of fuel rods that are placed on the first transfer section 210 at one time and are subjected to a helium leakage test. The lever 311 is rotated by the rotary driver. For example, the rotary driver may include a cylinder 314. A cylinder rod 314a driven linearly by the cylinder 314 is fastened to the lever 311 by a second hinge O. Thus, the lever 311 fastened to the cylinder rod 314a by the second hinge o is rotated about the hinge h by forward/backward movement of the cylinder rod 314a, and the first stopper 312 or the second stopper 313 protrudes upward from the transfer face. In the stopper member configured in this way, after the first stopper 312 protrudes, the fuel rods of the first transfer section 210 are stopped on the inclined face by the first stopper 312. When the second stopper 313 protrudes, the fuel rods located behind the second stopper 313 are prevented from moving downward by the second stopper 313, and the first stopper 312 is lowered downward. Thus, only the fuel rods located within the distance D are placed on the first transfer section 210, are loaded into the helium leakage testing chamber 100, and are subjected to the helium leakage test. When the first stopper 312 protrudes upward from the transfer face again, the fuel rods for the next helium leakage test move to the first stopper 312, and are on standby. Referring to FIG. 4, the lower transfer section 320 is located below the upper transfer section 310, and has the inclined face in the opposite direction of the inclined face of the upper transfer section 310. Thereby, the fuel rods 1 located on the lower transfer section 320 are rolled along the inclined face of the lower transfer section 320. Preferably, a testing table 400 is provided at a lower leading end of the lower transfer section 320. The fuel rods undergoing the helium leakage test are transferred to the testing table 400 via the lower transfer section 320. The fuel rods collected on the testing table 400 are subjected to macrography. The fuel rods undergoing the macrography are transferred to the next process via a manual conveyer 503, and are loaded on a fuel rod loading unit. The fuel rods loaded on the fuel rod loading unit are assembled into a fuel rod assembly as a nuclear fuel assembly. An operation of the fuel rod testing apparatus configured in this way will be described below. The manufactured fuel rods are guided and transferred for the helium leakage test and the macrography one by one by a conveyor 501 driven by an induction motor. The transferred fuel rods are transferred to the upper transfer section 310 of the main frame 300. The fuel rods transferred to the upper transfer section 310 move toward the fuel rod upward/downward transfer unit 200 along the inclination, and are stopped by the first stopper 312 that protrudes upward from the transfer face of the upper transfer section 310. Next, when the first stopper 312 is lowered, the fuel rods located between the first stopper 312 and the second stopper 313 are placed on the first transfer section 210. The gate of the helium leakage testing chamber 100 is opened, and then the drive roller 211 of the first transfer section 210 is driven. Thereby, the fuel rods are loaded into the helium leakage testing chamber 100. Each fuel rod may be marked with a bar code in which various pieces of information about a manufactured date, a type (enriched level) and so on are recorded for the purpose of managing the manufactured fuel rod. A bar code reader may be additionally installed on a side of the gate of the helium leakage testing chamber 100 so as to be able to read the bar code of the fuel rod to be tested in the process of loading the fuel rods into the helium leakage testing chamber 100. After the fuel rods are completely loaded into the helium leakage testing chamber 100, the gate is closed. The helium leakage test is performed on the fuel rods in the helium leakage testing chamber 100. After the test is completed, the gate of the helium leakage testing chamber 100 is opened. After the gate of the helium leakage testing chamber 100 is opened, the conveyer installed in the helium leakage testing chamber 100 is driven, and the second transfer section 220 of the fuel rod upward/downward transfer unit 200 is raised. The unloaded fuel rods are placed on the second transfer section 220 by driving of the drive roller of the second transfer section 220. After the unloaded fuel rods are completely placed on the second transfer section 220, the second transfer section 220 is lowered below a height of the lower transfer section 320, and the fuel rods are transferred to and placed on the lower transfer section 320. The fuel rods transferred to the lower transfer section 320 move to the testing table 400 along the inclined face of the lower transfer section 320. The fuel rods moving to the testing table 400 are subjected to the macrography, and then are transferred to the manual conveyer 503 by a given unit, and are transferred to the next process. In this way, in the fuel rod testing apparatus, the fuel rods loaded/unloaded into/from the helium leakage testing chamber are tested in a three-dimensional space making efficient use of upper and lower spaces of the main frame. Thereby, the helium leakage test and the macrography are effectively performed, and the space for facilities required for the processes can be efficiently used. Although an embodiment of the present invention has been described for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims. |
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claims | 1. A method for peening an obstructed region of a metal assembly that is partially obstructed by an obstructing part of the metal assembly comprising:determining an optimal peening path for treating the obstructed region irrespective of the obstructing part, said obstructing part being interposed between the optimal peening path and the obstructed region, the optimal peening path being only partially obstructed by said obstructing part;identifying a portion of the obstructing part within the optimal peening path, the portion of the obstructing part being arranged at a distance from the obstructed region such that a free space is delimited between the obstructed region and the obstructing part, the optimal peening path passing through the free space;determining a section of the portion of the obstructing part that is removable without affecting a mechanical integrity and functionality of the obstructing part;removing, by machining, the section so as to create additional space along the optimal peening path; andpeening the obstructed region, a path of the peening at least partially crossing through the additional space and through the free space. 2. The method as recited in claim 1 wherein the metal assembly is nuclear reactor pressure vessel. 3. The method as recited in claim 2 wherein the obstructing part penetrates a sloped wall of the nuclear reactor pressure vessel, a low hill side of the obstructing part defining an acute angle with the sloped wall, a high hill side of the obstructing part defining an obtuse angle with the sloped wall, the obstructed region being on a sloped wall on the low hill side of the obstructing part. 4. The method as recited in claim 3 wherein the obstructing part includes a radially enlarged section, the section being at the low hill side of the obstructing part, the radially enlarged section forming a free end of the obstructing part. 5. The method as recited in claim 3 the obstructing part is held in the sloped wall by a weld, the peening including peening the weld. 6. The method as recited in claim 5 wherein the peening of the weld including peening a surface of the weld at an angle of 10 to 60 degrees. 7. The method as recited in claim 5 wherein the weld is a J-weld. 8. The method as recited in claim 4 wherein the obstructing part includes a tubular section passing through the sloped wall, the radially enlarged section being a guide funnel fixed to a lower end of the tubular section. 9. The method as recited in claim 3 wherein the sloped wall is hemispherical. 10. The method as recited in claim 9 wherein the sloped wall is part of a closure head of the nuclear reactor pressure vessel. 11. The method as recited in claim 10 wherein the obstructing part is a core exit thermocouple nozzle. 12. The method as recited in claim 8 wherein the machining forms a notch in a top edge of the guide funnel delimiting the additional space, the path of the peening at least partially crossing through the notch. 13. The method as recited in claim 12 wherein the notch extends at a circumferential angle in a range of 60 degrees to 180 degrees. 14. The method as recited in claim 8 wherein the guide funnel covers the lower end of the tubular section before the machining, the machining uncovering a section of the lower end. 15. The method as recited in claim 1 wherein the machining is electrical discharge machining. 16. The method as recited in claim 1 wherein the peening is cavitation peening or laser peening. 17. The method as recited in claim 4 wherein the peening is cavitation peening, and wherein the radially enlarged section of the obstructing part is surrounded by a container of liquid during the cavitation peening. 18. A method for peening an obstructed region of a metal assembly that is obstructed by an obstructing part of the metal assembly, the metal assembly being a nuclear reactor pressure vessel, comprising:determining an optimal peening path for treating the obstructed region irrespective of the obstructing part;identifying a portion of the obstructing part within the optimal peening path;determining a section of the portion of the obstructing part that is removable without affecting a mechanical integrity and functionality of the obstructing part;removing, by machining, the section so as to create additional space along the optimal peening path; andpeening the obstructed region, a path of the peening at least partially crossing through the additional space;wherein the obstructing part penetrates a sloped wall of the nuclear reactor pressure vessel, a low hill side of the obstructing part defining an acute angle with the sloped wall, a high hill side of the obstructing part defining an obtuse angle with the sloped wall, the obstructed region being on a sloped wall on the low hill side of the obstructing part;wherein the sloped wall is hemispherical and is part of a closure head of the nuclear reactor pressure vessel;wherein the obstructing part is a core exit thermocouple nozzle. 19. A method for peening an obstructed region of a metal assembly that is obstructed by an obstructing part of the metal assembly, the metal assembly being a nuclear reactor pressure vessel, comprising:determining an optimal peening path for treating the obstructed region irrespective of the obstructing part;identifying a portion of the obstructing part within the optimal peening path;determining a section of the portion of the obstructing part that is removable without affecting a mechanical integrity and functionality of the obstructing part;removing, by machining, the section so as to create additional space along the optimal peening path; andpeening the obstructed region, a path of the peening at least partially crossing through the additional space;wherein the obstructing part penetrates a sloped wall of the nuclear reactor pressure vessel, a low hill side of the obstructing part defining an acute angle with the sloped wall, a high hill side of the obstructing part defining an obtuse angle with the sloped wall, the obstructed region being on a sloped wall on the low hill side of the obstructing part;wherein the obstructing part includes a radially enlarged section, the section being the low hill side of the obstructing part, the radially enlarged section forming a free end of the obstructing part;wherein the obstructing part includes a tubular section passing through the sloped wall, the radially enlarged section being a guide funnel fixed to a lower end of the tubular section;wherein the machining forms a notch in a top edge of the guide funnel delimiting the additional space, the path of the peening at least partially crossing through the notch, and/or the guide funnel covers the lower end of the tubular section before the machining, the machining uncovering a section of the lower end. |
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description | This is a non-provisional application under 35 U.S.C. §1.111(a) which claims priority from Korean patent application 10-2007-0086024 filed on Aug. 27, 2007, which is incorporated herein by reference. 1. Field of the Invention The present invention relates to a spacer grid used for placing and supporting a plurality of nuclear fuel rods within a nuclear fuel assembly, and more particularly, to an anti-fretting wear spacer grid with canoe-shaped springs capable of preventing the fretting wear of fuel rods, thereby avoiding the damage of the fuel rods. 2. Background of the Related Art A nuclear reactor is a device that artificially controls the chain reaction of the nuclear fission of fissile materials, thereby achieving a variety of use purposes such as the generation of heat, the production of radioisotopes and plutonium, the formation of radiation fields, or the like. Generally, enriched uranium that is obtained by raising a ratio of uranium-235 to a range between 2% and 5% is used in a light water nuclear reactor. The uranium is molded to a cylindrical pellet having a weight of 5 g and is processed to a nuclear fuel used in the nuclear reactor. Numerous pellets are piled up to form hundreds of pellet bundles and then put into a cladding tube made of Zircaloy being at a vacuum state. After that, a spring and a helium gas are put thereinto, and an upper end plug is welded thereon, thereby making a fuel rod. The fuel rod is finally surrounded by a nuclear fuel assembly and then burnt up within the nuclear reactor through nuclear reaction. The nuclear fuel assembly and the parts therein are shown in FIGS. 1 and 2. FIG. 1 is a schematic view showing a general nuclear fuel assembly, FIG. 2 is a top plane view showing the spacer grid, and FIG. 3 is a cut-off perspective view showing the spacer grid. Referring to FIG. 1, the nuclear fuel assembly basically includes a frame body comprised of a top nozzle 4, a bottom nozzle 5, guide thimbles 3, and a plurality of spacer grids 2, and a plurality of fuel rods 1 held longitudinally in an organized array by the spacer grids 2 spaced along the length thereof in such a manner as to be supported by means of springs 6 (see FIGS. 2 and 3) and dimples 7 (see FIGS. 2 and 3) disposed within the spacer grids 2. So as to prevent the formation of the scratches on the fuel rods 1 and the generation of the damage on the springs within the spacer grids 2 upon assembling the nuclear fuel assembly, thereafter, the fuel rods 1 have a locker applied thereon and are then inserted longitudinally into the frame body of the nuclear fuel assembly. Next, the top and bottom nozzles are secured to the opposite ends of the nuclear fuel assembly, thereby finishing the assembling procedure of the nuclear fuel assembly. Then, after the locker of the finished assembly is removed, the distances between the fuel rods 1, the distances between the fuel rods 1, the distortion of the nuclear fuel assembly, the total length thereof, and the dimension thereof are checked out, thereby finishing the manufacturing procedure of the nuclear fuel assembly. Referring to FIGS. 2 and 3, the spacer grid is made in a lattice by coupling a plurality of slots (not shown) formed by a plurality of strips (thin metal plate) connected with one another so as to define a plurality of space portions into which the fuel rods 1 are held longitudinally thereby. About 10 to 13 spacer grids are arranged along the length direction of the nuclear fuel assembly and welded to the guide thimbles 3 having a length of 4 m. The springs 6 and the dimples 7 are regularly formed on each space portion defined by the spacer grid 2, such that as they are brought into contact with the fuel rod 1 (see FIG. 1), the distance between the fuel rods 1 is maintained and arranged at their defined position. Further, the fuel rods 1 are fixed by the elasticity of the springs 6. On the other hand, the recent development of the nuclear fuel is aimed to obtain high burn-up performance and integrity. So as to develop high burn-up fuel, there are a variety of methods of enhancing the thermal transmission efficiencies from the fuel rods to coolant. Many of the methods are introduced wherein the improvement in the flow characteristics of coolant around the fuel rods is effectively accomplished by attachment of mixing blades to the spacer grid, a change of the shape of the mixing blades, or an appropriate configuration of coolant flow channels in the spacer grid. However, the above-mentioned methods of enhancing the thermal transmission efficiencies of nuclear fuel assemblies also generate turbulences in coolant flowing around the fuel rods, and the turbulences of the coolant undesirably cause flow-induced vibration by which the fuel rods are vibrated. The flow-induced vibration of the fuel rods 1 causes the fuel rods to slide against their contact surfaces at which the fuel rods are brought into contact with the springs and dimples of the spacer grids, such that the contact surfaces are partially abraded to cause fretting wear of fuel rods to occur, which results in the damage on the fuel rods. That is, the above-mentioned methods for improving the thermal performance so as to develop the high burn-up fuel result in the acceleration of the damage of the fuel rods. As the burn-up is conducted, on the other hand, the irradiation growth of the spacer grid is performed transversely. Also, the fuel rods are repeatedly contracted by the radial creeps in the burn-up process in the reactor, that is, by the high pressure caused by the coolant in the reactor, and extended radially by the expansion of fuel pellets, such that the outer diameters of the fuel rods have irregular directionality, which generates the gap between the springs/dimples of the spacer grid and the fuel rod, thereby causing much fretting wear. So as to reduce the fretting wear, the contact length between the fuel rod and the springs/dimples is extended longitudinally, and otherwise, the surface-contact therebetween is generated, so that even though the fretting wear occurs, the wear depth is substantially reduced even under the same wear area. FIGS. 4a and 4b show the conventional spacer grid having a spring shape generally used, wherein the linear contact length between the fuel rod and the springs/dimples in the spacer grid is extended longitudinally. The springs protrude horizontally and vertically on a grid surface of the strip and support a nuclear fuel rod. In the conventional spacer grid, each spring is flat-shaped in a manner of linear contact with the fuel rod, so that the linear contact manner more effectively protects the fuel rod from fretting wear. However, actually, the spring is irregular on the flat surface thereof to obtain an elastic force when supports the fuel rod, as shown in FIG. 4a. Therefore, the initial contact between the fuel rod and the spacer grid are not linear contact, but are three-point contact. As the fretting wear is developed, the three-point contact is changed into the linear contact. At this time, the fretting wear of the fuel rod is accelerated due to the initial three-point contact. On the other hand, there has been proposed a method for reducing the fretting wear by the generation of the surface contact of the spacer grid with the fuel rod, which is disclosed U.S. Pat. No. 6,606,369 (hereinafter, referred to as ‘prior art’) entitled ‘Nuclear reactor with improved grid, as filed Mar. 6, 2002. According to the prior art, as shown in FIGS. 5a and 5b, a spring 62 has a curved surface in such a manner as to have the surface contact with the fuel rod 1, which more effectively prevents the movements of the fuel rod in axial and transverse directions caused by the flow-induced vibration, in comparison with the conventional point or linear contact manner. According to the prior art, however, the contact surface of the spring 62 and the dimples 72 with the fuel rod 1 is formed in the same length as the curvature radius of the fuel rod 1, which really makes it impossible to form or maintain the surface contact in an accurate manner by the manufacturing tolerance of the spacer grid and by the variation of the roundness of the fuel rod 1 and also to maintain constant roundness and curvature radius during the burn-up process of the fuel rod. Further, in case where the contact portion of the spring does not have a theoretically complete curved surface, the irregular linear or point contact of the spring 62 with the fuel rod 1 is formed to cause unexpected fretting wear to happen. Accordingly, the present invention has been made in view of the above-mentioned problems occurring in the prior art, and it is an object of the present invention to provide an anti-fretting wear spacer grid with canoe-shaped springs that can provide the linear contact with a fuel rod such that the linear contact is extended as long as possible from initial contact to final contact, thereby achieving greater reduction of the fretting wear depth when compared with the same fretting wear area. It is another object of the present invention to provide an anti-fretting wear spacer grid with canoe-shaped springs wherein the shape of the contact surface between the canoe-shaped springs and a fuel rod can be constantly maintained through the entire manufacturing procedure of the fuel rod in the nuclear reactor. To accomplish the above objects, according to the present invention, there is provided an anti-fretting wear spacer grid having a plurality of canoe-shaped springs, the spacer grid including: a plurality of strips adapted to define a plurality of unit grid cells; a pair of upper and lower dimples formed protrudedly in longitudinal and transverse directions on one surface of each unit grid cell; and each of the canoe-shaped springs formed between the pair of upper and lower dimples on one surface of each unit grid cell in such a manner as to be projected to an opposite direction to the projected direction of the dimples, each of the canoe-shaped springs having a given elastic force so as to support the fuel rod together with the dimples, wherein each of the canoe-shaped springs includes: a fuel rod-contacting part having a flat surface having a predetermined longitudinal length so as to have linear contact with the fuel rod; a curved face-connecting part formed on the upper and lower portions of the fuel rod-contacting part; a leg-connecting part formed on the end portion of the curved face-connecting part; and legs each being formed of a plate shape having a predetermined length and connecting the both sides of the leg-connecting part with one surface of each unit grid cell such that the leg-connecting part is projected to a predetermined height from one surface of each unit grid cell, each of the legs being adapted to apply a given elastic force to the canoe-shaped spring. According to the present invention, therefore, there is provided an anti-fretting wear spacer grid with canoe-shaped springs that can provide the linear contact with a fuel rod such that the linear contact is extended as long as possible from initial contact to final contact, thereby achieving greater reduction of the fretting wear depth when compared with the same fretting wear area. Additionally, there is provided an anti-fretting wear spacer grid with canoe-shaped springs wherein the shape of the contact surface between the canoe-shaped springs and a fuel rod can be constantly maintained through the entire manufacturing procedure of the fuel rod in the nuclear reactor, thereby diminishing the partial abrasion of the fuel rod due to irregular contact. Hereinafter, an explanation on an anti-fretting wear spacer grid with canoe-shaped springs according to the present invention will be given with reference to the attached drawings. In the following description, wherever possible, the same reference numerals will be used throughout the drawings and the description to refer to the same or like parts. On the other hand, the individual space portion defined by the lattice structure of the spacer grid is called a grid cell, and one surface of the interior of the grid cell is called one grid cell surface. Also, if one grid cell surface is disposed in a longitudinal direction, all of the grid cell surfaces located parallel thereto become the grid cell surface being in the longitudinal direction, and the grid cell surfaces located perpendicular to the grid cell surface being in the longitudinal direction becomes those being in the transverse direction. Further, an axial direction is a length direction of the unit grid cell, to which the fuel rod is inserted into the unit grid cell. As shown in FIGS. 6a and 6e, the canoe-shaped spring 63 disposed on every grid cell surface of the spacer grid is a principal part in the preferred embodiment of the present invention. FIG. 6a is a front view showing one unit grid cell surface having the canoe-shaped spring 63 formed thereon, FIG. 6b shows the side surface of one unit grid cell surface, FIG. 6c is a perspective view showing one unit grid cell surface so as to make the canoe-shaped spring 63 well understood, FIG. 6d is a sectional view showing the portion P-P′ of one unit grid cell surface, and FIG. 6e is a top plane view showing the unit grid cell having the canoe-shaped springs formed thereon, into which the fuel rod is imaginarily put. As shown in FIG. 6a, the canoe-shaped spring 63 basically includes a fuel rod-contacting part 631, a curved surface-connecting part 632, a leg-connecting part 633, and legs 634. Referring to FIGS. 6a and 6d, first, an explanation of the fuel rod-contacting part 631 will be given. The fuel rod-contacting part 631 has a predetermined length in the longitudinal direction thereof, as shown in FIG. 6a, and is flat along the middle portion thereof so as to have the linear contact with the fuel rod and is bent toward the spacer grid at the both side portions thereof so as to support the flat middle portion of the fuel rod-contacting part 63, as shown in FIG. 6d, thereby preventing the fuel rod-contacting part 631 from being bent upon the insertion of the fuel rod and during the activation of the nuclear reactor. Referring to FIGS. 6a and 6b, next, an explanation of the curved surface-connecting part 632 will be given. The curved surface-connecting part 632 is located between the fuel rod-contacting part 631 and the leg-connecting part 633 as will be discussed later, on the upper and lower portions of the fuel rod-contacting part 631. The curved surface-connecting part 632 has the same outer diameter and shape as the fuel rod-contacting part 631 at the contact portions with the fuel rod-contacting part 631 and has the same outer diameter and contour as the leg-connecting part 633 at the contact portions with the leg-connecting part 633. That is, the curved surface-connecting part 632 has a gradually reduced outer diameter toward the leg-connecting part 633 from the fuel rod-contacting part 631. The curved surface-connecting part 632 serves to connect the fuel rod-contacting part 631 and the leg-connecting part 633 having different outer diameters and contours from each other. The connected portions of the curved surface-connecting part 632 with the fuel rod-contacting part 631 and the leg-connecting part 633 are desirably formed of a gentle curved surface, which prevents the fuel rod from being damaged upon the insertion of the fuel rod into the grid cell. Referring to FIGS. 6a and 6c, the leg-connecting part 633 will be discussed below. The leg-connecting part 633 is located at each end portion of the curved surface-connecting parts 632. The leg-connecting part 633 serves to connect the legs 634 as will be discussed later therewith and has a semi-cylindrical shape being hollow at the center thereof. In case where the leg-connecting part 633 is connected to the legs 634, as shown in FIG. 6b, it projects to a predetermined height from one grid cell surface. On the other hand, the leg-connecting part 633 desirably has a smaller outer diameter than that of the fuel rod-contacting part 631 so as to sufficiently form a space portion into which the legs 634 are disposed. Referring to FIGS. 6a and 6c, an explanation of the legs 634 will be given. Each of the legs 634 is of a flat shape having a predetermined length. The legs 634 are connected to the grid cell surface from the both sides of the leg-connecting part 633. At this time, the conventional legs are generally formed perpendicular to the direction of the insertion of the fuel rod, that is, to the axial direction, but the legs 634 of the present invention are desirably formed oblique with respect to the axial direction so as to permit the stiffness of the spring to be sufficiently lowered. On the other hand, the dimples 7 have the same structure as well known in the art, and therefore, an explanation on them will be avoided for the brevity of the description. Hereinafter, an explanation on the operations of the anti-fretting wear spacer grid with the canoe-shaped springs according to the present invention will be given. In case where the fuel rod (not shown) is inserted longitudinally into the unit grid cell on which the springs 63 are formed, as shown in FIG. 6e, the fuel rod is brought into contact with the fuel rod-contacting part 631. At this time, the bending portions of the fuel rod-contacting part 631 formed at the both side portions thereof function to maintain the linear contact shape between the fuel rod and the fuel rod-contacting part 631 during the entire life time of the nuclear fuel assembly, thereby enhancing the stiffness of the fuel rod-contacting part 631 and preventing the shape of the fuel rod-contacting part from being distorted. At this time, the legs 634 function to support the fuel rod-contacting part 631 and to provide a given elastic force between one unit grid cell surface and the fuel rod-contacting part 631. In this case, the longer the legs 634 are, the lower the stiffness of the spring 63 becomes. In the preferred embodiment of the present invention, the leg-connecting part 633 has a relatively small outer diameter, thereby ensuring the space portion into which the legs 634 are disposed, and further, the legs 634 are formed oblique with respect to the axial direction, thereby permitting the legs 634 to be extended longer. That is, in the preferred embodiment of the present invention, the leg-connecting part 633 having the smaller outer diameter and the obliquely formed legs 634 provide good flexibility to the spring 63. The two canoe-shaped springs 63 and the four dimples 7 are formed within each unit grid cell of the spacer grid, so that the linear contacts with the fuel rod are formed on the total six contact portions thereof, thereby restricting the movement of the fuel rod. While the present invention has been described with reference to the particular illustrative embodiments, it is not to be restricted by the embodiments but only by the appended claims. It is to be appreciated that those skilled in the art can change or modify the embodiments without departing from the scope and spirit of the present invention. |
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claims | 1. A method of separating osmium from iridium comprising:receiving a powdered mixture of osmium and iridium;oxidizing the osmium of the powdered mixture;separating the oxidized osmium into different isotopes of osmium, including osmium 191;capturing the oxidized osmium 191 a trapping solution;reducing the oxidized osmium 191 solution to release the osmium 191. 2. The method according to claim 1, wherein the oxidizing comprises placing the powdered mixture in a solution and heating the solution to transfer the oxidized osmium to gaseous form. 3. The method according claim 1, wherein the trapping solution is heated to about 110° C. 4. The method according to claim 1, wherein the osmium is oxidized using chromium trioxide (CrO3) in a sulfuric acid solution. 5. The method according to claim 1, wherein the osmium in the powder is oxidized by the following equation:3Os +4H2Cr2O7+12H2SO4→3OsO4↑+4 Cr2(SO4)3+16H2O. 6. The method according to claim 1, wherein the osmium is oxidized using a potassium permanganate (KMnO4) in a sulfuric acid solution. 7. The method according to claim 6, wherein the osmium in the powder is oxidized by the following equation:8KMnO4+12H2SO4+5Os =5OsO4+8MnSO4=4K2SO4+12H2O. 8. The method according to claim 7. wherein the method includes a side reaction having the following equation:4KMnO4+2 H2SO4=4MnO2+2 K2SO4+3 O2+2H2O. 9. The method according to claim 1, wherein the oxidized osmium is agitated by air or N2 passing through a system for performing the method. 10. The method according to claim 1, wherein the oxidized osmium 191 is trapped in a KOH solution in one or more capturing stages. 11. The method according to claim 10, wherein the oxidized osmium 191 of the powdered mixture is trapped in the KOH solution forming a complex according the to following formula:OsO4+2KOH →K2 [OsO4 (OH)2]. 12. The method according to claim 10, wherein the KOH solution has between 10% to 25% KOH w/v. 13. The method according to claim 10, wherein a sodium hydrosulfide solution is added to the KOH solution to precipitate the osmium as osmium disulfide (OsS2). 14. The method according to claim 13, wherein the osmium is precipitated according to the following equations:NaHS +2KOH (Excess) →K2S +H2O +NaOH2K2[OsO4 (OH)2]+5K2S +4H2O→2OsS2↓+12KOH +K2SO4. 15. The method according to claim 1, wherein said separating is performed by laser isotope separation. 16. The method according to claim 1, wherein said separating is performed by electromagnetic isotope separation. 17. The method according to claim 1, wherein said separating is performed by diffusion isotope separation. 18. The method according to claim 1, wherein said separating is performed by SILEX isotope separation. 19. The method according to claim 1, wherein said separating is performed by centrifugal isotope separation. |
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058752200 | claims | 1. A process for the production of radiostrontium, said process comprising the following steps: bombarding a target of metallic rubidium by a beam of accelerating charged particles, melting said irradiated target of metallic rubidium, immersing a sorbing material into said melt of metallic rubidium, extracting radiostrontium by sorption on the surface of said sorbing material, and using, as said sorbing material, a material selected from the group consisting of heat-resistant metals, metallic and silicon oxides, said material being inert with respect to said rubidium, and extracting the resultant radiostrontium. the temperature of molten rubidium is set to be close to the optimum one for the desorption of radiostrontium within the range of from 220 deg.C. to 270 deg.C. 2. The process as claimed in claim 1, wherein the temperature of said sorbing material is set to be close to the optimum one for the sorption of radiostrontium within the range of from the melting point of metallic rubidium to 220 deg.C., and |
040101088 | summary | BACKGROUND OF THE INVENTION The present invention relates to improvements in RADIOACTIVE WASTE DISPOSAL, and more particularly to the disposal of radioactive materials by immobilizing them within a solid mass for storage and/or burial. It is well known that waste products occur as a natural result of activity involving the use of radioactive isotopes. For example, waste products are provided during the operation of atomic reactors and the like, and these waste products may be produced directly from primary radiation sources or secondarily by the creation of isotopes from non-radioactive metals or the like. In order to assure smooth efficient continuation of atomic processes generating such waste material, efficient disposal means must be provided both for primary and secondary waste products. At the present time, disposal has been achieved by immobilizing the waste in a solid block, and then by disposal at sea or by burial in a specially designated burial site. Burial at sea requires more and more preparation, because of the long range effects of certain pollution components that might build up. When the product is disposed of at a burial site, it is also necessary to provide safe means for transporting the material to the burial site. In addition, it is important to assure the containment and safe storage of the material at the burial site for a time sufficient to allow a sufficent decay of the radioactive components to reduce the radiation intensity thereof to a relatively safe level. Thus it is seen that whatever the disposal of the waste material, it is important to provide means for protecting the material and assuring its safe storage at the disposal site for a long period of time. Prior to this invention, Portland Cement has been in rather widespread use for the purpose of encapsulating and holding radioactive waste material therewithin so as to provide a protective block for the material at the burial site. Portland cement has been found to be particularly advantageous where the radioactive waste material is present in water, and it is advantageous to dispose of a certain amount of water along with the radioactive waste material in order to provide an efficient handling process. For example, the water utilized in the cooling loop of atomic reactors tends to accumulate contaminations of radioactive nickel and cobalt probably as a result of conversion of iron and/or nickel in the tubes carrying the water. In any event, these materials build up in the water so that it is important to remove the waste from time to time in order to prevent a buildup from reaching a very hot or hazardous level. In such a case, probably the most serious component is cobalt 60, because it emits hard gamma rays and has a half-life of approximately five years. Prior to this invention, the cooling water was removed and mixed with Portland Cement in the usual water-cement ratio, allowed to solidify and then the block of cement buried in a waste dump. Such disposal has been generally satisfactory for many operations, but it has a number of disadvantages. One of the disadvantages resides in the heavy weight of the cement and the like, which must be transported often over a considerable distance. Another disadvantage, and perhaps a more serious one, resides in the fact that many waste products of this general class now contain levels of boron material that render disposal in Portland Cement unsatisfactory or impossible because of the lack of compatability of the materials. Other areas of improvement are also seen to be available, such as the handling problems occuring with cement in processing equipment and the possibility of the cement setting up in an undesired fashion during an unexpected shutdown. Rather than go into all of the disadvantages of the cement process, it is proposed to provide an improved process in which certain advantages are achieved, and which is particularly suitable for disposing of waste products having high concentrations of compounds containing boron. Another problem which has been of some concern with the use of Portland Cement is the possibility of the radioactive material leaching therefrom. This problem is particularly acute where disposal at sea is contemplated, and efforts to utilize materials other than Portland Cement have generally been in the area of the use of hydrophobic materials so as to render the solid block substantially leach-proof. However, the use of hydrophobic materials such as bitumen or asphalt has a number of disadvantages particularly in the mixing and processing steps, and the use of these materials has generally been rejected as not substantially improving the situation involved with the use of Portland Cement. SUMMARY OF THE INVENTION From the above background material, it is seen that a primary object of the present invention is to provide a process for making a disposable waste product material in which improvements are made over the use of Portland Cement in order to increase the range of disposable materials, provide reduction of weight required for shipping, and generally provide a more reliable disposal from the standpoint of safety and the like. These and other objects are achieved by solidifying the wet or water-carried waste product through the steps of adding a hydrophilic resinous material to the waste in an amount sufficient to set up and cure into a solid block, mixing the materials together to provide the desired distribution of waste materials therein, and curing the material to a solid mass. In general, it is believed that any hydrophilic resinous material capable of taking up water upon curing will be suitable to render the wetted or water-carried waste material immobile and shielded therewithin. However, the preferred hydrophilic resin is any of the usual urea-formaldehyde compositions, which are available commercially in the partially polymerized state, and capable of curing to a high polymer upon the addition of an acidic curing agent. After the radioactive waste material is thus immobilized within a solid block of hydrophilic resinous material, it may be waterproofed to protect against leaching, if desired. This objective may be achieved by the addition of a substantially waterproof resin as a coating thereover, or by a cover or any other protective waterproofing material that will prevent transfer of water from within to the outside and reverse. Another object of the present invention is to provide improvements within this general process of providing a safe immobilized waste product, and to increase the efficiency of the use of materials and the like used up in the process. Thus in the preferred form of the invention, the radioactive waste material is first concentrated to a level more suitable for disposal, but still at a sufficiently low level so as to remain within the low hazard classifications. Where the radioactive material is present in water, this concentration is obtained by water removal. In the case of removal of radioactive waste from the water in the cooling loop of a reactor, the removed water may advantageously be returned to usage for further cooling. In such a case, water containing radioactive ions such as radioactive iron, nickel, and cobalt, are brought in contact with ion exchange resin beads capable of taking up such cations and holding them within the resin mass. The water which is thus deionized and thereby has its radioactive metallic ion component substantially removed is returned to the cooling loop, and the wet resin beads containing the radioactive components are then disposed of by encapsulating them within a hydrophilic resinous material as explained above, In general, any ion exchange resin capable of picking up radioactive waste components may be used. However, where it is desired to remove iron, nickel and cobalt ions, cation exchangers should be used. Cation exchange resins are well known, and available commercially. A typical ion exchange resin preferred in the practice of the process of this invention has a styrene-divinylbenzene matrix which is suitably sulfonated to provide a strongly acidic, cation exchange resin in the form of beads. Such resins are sufficiently dense and insoluble in water to provide easy separation, yet are sufficiently hydrophilic to provide the desired ion exchange activity as well as to provide compatability with the hydrophilic resins utilized in accordance with the present invention. It will be appreciated that absorbing agents in general, which may or may not be classified as ion exchange resins, but which are capable of picking up the desired radioactive component are also suitable. In this connection, materials such as diatomaceous earth, Powdex (powdered filter aid) Solco Foc (wood cellulose flour) and the like are suitable. In such case, the substances may be filtered out advantageously to provide solids having concentrates of wastes therein. When a typical ion exchange resin is used, instead of filtering same, the resin may be regenerated after separation in a more concentrated solution and the regenerated resin beads recycled for reuse. Another method of concentrating the materials is simply by vacuum evaporation of the water, and the water vapor may be condensed and returned again to the process from whence it came, if desired. While it will be seen that any of these methods for concentrating the waste materials may be suitable in and of themselves, it is also sometimes advantageous to provide a combination of methods so as to provide a controlled concentration of waste and water in proper proportion for mixture with the resin. In addition, filter aids and filtration may be utilized instead of ion exchange beads to concentrate the materials and locate them in certain desired areas within the final solid resinous block. It is also desirable to add filler material or the like to extend the resin and also act as an additional shield for the radioactive components. In other words, the solids of this invention not only hold and immobilize the waste material, but they provide a primary shield therefor so that the radiation such as hard gamma rays are reduced before leaving the solid mass in which the radioactive sources are contained. It will also be appreciated that any other suitable filler material may be added to the resinous components in accordance with those materials suggested in the literature for use with the particular resin involved. In all such cases, the amount of filler will be determined by conventional standards, i.e. the amount which will best extend and increase the use of the resin itself, but will stay within the ranges of physical properties desired for the final composition. The use of hydrophilic resins in accordance with the present invention is particularly advantageous with regard to handling of water solutions and wet materials. Such handling not only has the advantage of allowing water to be utilized as a carrier for pumping and other handling and the like, but it also provides the build-up advantages of having water present during the exothermic polymerization reaction. During polymerization, the high heat capacity of the water prevents undue heat built up and provides for proper curing without thermal breakdown. In addition, it provides a convenient method for getting rid of water that may contain waste in and of itself either as a primary carrier, or as a cleaner utilized to flush out radioactive material from the system. It has also been found that Portland Cement and hydrophilic resinous materials do not hold the water and associated ions in a sufficiently strong bond for certain disposal applications, such as disposal at sea. In such cases, it is contemplated that the solid mass will be further encapsulated in one or more waterproof materials. For example, the solid waste block may be advantageously prepared in a metal container such as a drum and the metal container disposed of along with the resin and waste product. In such a case, however, the metal container may disintegrate or corrode away and expose the resin block too quickly, particularly when subjected to corrosive action of sea water. Accordingly, it is preferred to coat and capsulate or otherwise cover the hydrophilic resin block containing the waste material therein. In other words, a substantially waterproof or water impervious resinous material in the form of a coating or a bag or any other device that will assure containment may be used. If desired, such further material may be carried in a metal container. For example, the process of this invention may be practiced by utilizing a large metal container such as a drum, lining the container with a polyethylene bag material, with the sides extending sufficiently to provide a fold-over enclosure. With the procedure, the radioactive waste material, resin components, and any other of the materials suggested for use in accordance with the process of this invention are then added, and the resin cured to provide a solid block within the plastic bag and held within the container. The bag is then folded over the top and sealed to provide a waterproof coating, and the metal container is then closed. Where such a container is used, leaching of the waste materials will not occur even after the metal container has corroded away. Alternative to the bag process, it may be advantageous to utilize resins that will adhere to the hydrophilic resin utilized in the primary process. Such processes may be carried out by first curing a base lining in the bottom of the container, then curing the plastic mass within the container, with curing providing a certain amount of shrinkage, and then curing the waterproof or water repellent resin in the further stage around the side and top so as to completely fill the container and provide a water resistant protective layer. For example, when the preferred urea-formaldehyde resin is used for solidifying and retaining the radioactive waste material in accordance with this invention, the water resistant material may be a butylated urea-formaldehyde or a melamine-formaldehyde resin. These resins have improved resistance to a leaching effect of water. Alternatively, a typical hydrophobic resinous material may be utilized instead of, but in the same manner, by using an asphaltic or bituminous material first as a layer on the bottom and then to fill the side and top voids after processing and shrinking. Further alternatives and advantages of the invention will become apparent as the specification progresses and the new and useful features of the radioactive disposal described herein will be more fully defined in the claims attached hereto. DESCRIPTION OF THE PREFERRED EMBODIMENTS The preferred hydrophilic resin to be used in accordance with this invention is any of the urea-formaldehyde resins available from a plurality of commercial sources as standard articles of commerce. These resins are prepared by reacting urea and formaldehyde in mol ratios between about 1:1 and 1:4 respectively, and preferably between 1:1.5 and 1:2.5. For optimum results, the mol ratio is about 1 part urea to about 2 parts formaldehyde. Typically, solid urea and an aqueous solution of formaldehyde are reacted with one another to produce a resin syrup that is in the thermosetting state but capable of being converted to a thermoset state. These resins are available in syrup form, and sometimes available in a spray-dried form, which may be redispersed in water to a desired solids content. Since part of the water will come from the waste material, the urea-formaldehyde should be in a concentrated form with the final ratio of resin solids and water being present in the final dispersion in a ratio of about 21/2 to about 5 parts water per part resin by weight and preferably from about three to about 4 parts water per part resin solids. A typical catalytic material used to convert the urea resin to a thermoset state at ambient temperature is an acidic material having a dissociation constant between about 10.sup.0 to 10.sup.-.sup.5. The amount of catalytic material used will depend upon the strength of the acidic material used and upon the nature of the composition in which it is used. For example, materials like boric acid tend to inhibit the polymerization, and therefore increased catalyst is required to achieve the same cure time. However, generally the amount of acidic catalytic material will be between say about 0.3 and 20% by weight of the resin solids in the mixture. In general, any acid capable of providing a pH below 5 in the dispersion may be utilized, as is well known in the art, and it is preferred to utilize sodium bisulfate, since it is available as a solid and provides an excellent strength acid. Certain materials such as filter aids, ion exchange resins and materials that act as one of these or both are usually added in order to improve processing and provide the most economical and practical way to eliminate waste. However, any of these materials which are compatible with the urea-formaldehyde are suitable, and considerable latitude is permissible in this area. |
abstract | The object of the present invention is a method for establishing so-called “mixed IN-CORE mappings”. Its essential purpose is to compensate a loss of density of a reference instrumentation, called “RIC instrumentation” (or “RIC system”), when a significant number of locations, initially used by the sensors of the RIC system, are occupied by fixed collectron-type rods. An obvious physical interest lies in the increase of the measurement density, and thus of the level of confidence, associated with the operating results deduced from the processing of these measurements. One application of the method according to the invention concerns a collectron-type detector calibration method placed inside a nuclear reactor core. |
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claims | 1. An integrated head assembly for a commercial nuclear reactor having a plurality of control element drive mechanisms and a reactor vessel closure head, the integrated head assembly comprising:a lower portion including hardware for attaching the lower portion to the nuclear reactor vessel closure head;a separable upper portion that is mechanically uncoupled from the lower portion when the lower portion and upper portion are assembled and in use and further comprises means for supporting substantially the entire weight of the upper portion over the lower portion such that the lower portion does not support any substantial portion of the weight of the upper portion during operation of the reactor;a duct disposed externally of the upper portion and the lower portion, the duct extending from the upper portion to the lower portion; anda forced air cooling system attached to the upper portion and oriented to generate air flow about the control element drive mechanisms;wherein the separable upper portion and the lower portion are configured to be removable separately during refueling outages; andfurther wherein the upper portion comprises a plurality of upright supports having footpads such that during storage the upper portion can be set on a flat surface and supported thereon by the upright supports. 2. The integrated head assembly of claim 1, wherein the lower portion further comprises a ring beam that is adapted to engage lifting lugs on the reactor vessel closure head. 3. The integrated head assembly of claim 2, wherein the lower portion further comprises a plurality of lifting rods that are attachable to the reactor vessel closure head such that the lower portion of the integrated head assembly and the reactor vessel closure head can be lifted as a unit. 4. The integrated head assembly of claim 1, wherein the lower portion includes a cylindrical wall and a transverse plate, and wherein the transverse plate has a plurality of apertures adapted to accommodate the plurality of control element drive mechanisms. 5. The integrated head assembly of claim 1, wherein the forced air cooling system comprises a plurality of fans fluidly connected to an upper plenum, wherein the fans are adapted to draw air over the control element drive mechanisms. 6. The integrated head assembly of claim 5, wherein the forced air cooling system further comprising a chiller disposed in the upper plenum that is operable to cool air in the upper plenum. 7. The integrated head assembly of claim 6 wherein the chiller is attached to an external cold water source. 8. The integrated head assembly of claim 1, wherein the means for supporting the weight of the upper portion over the lower portion comprises a plurality of horizontal beams adapted to support the weight of the upper portion during use. 9. The integrated head assembly of claim 1, wherein the upright structural supports include lifting connectors adapted to be engaged by a tripod assembly to facilitate moving the upper portion. 10. The integrated head assembly of claim 1, further comprising a plate attached to the horizontal beams to define an access platform. 11. The integrated head assembly of claim 5, wherein the lower portion further comprises a lower plenum, and wherein the duct fluidly connects the lower plenum with the upper plenum. 12. The integrated head assembly of claim 11 wherein the duct comprises a lower end having a flexible joint that is releasably attachable to the lower plenum such that no significant loads are transmitted by the duct to the lower portion of the integrated head assembly. 13. An integrated head assembly for a nuclear reactor disposed in a containment between first and second containment walls, the integrated head assembly comprising:a lower portion including hardware for attachment to the nuclear reactor, the lower portion having an annular plenum;a separable upper portion disposed directly over the lower portion that is mechanically uncoupled from the lower portion when the lower portion and upper portion are assembled and in use such that the lower portion does not support any substantial portion of the weight of the upper portion during operation of the reactor, and having a plurality of outwardly extending support beams such that substantially the entire weight of the upper portion is supportable by the support beams, the upper portion further including at least one fan fluidly connected to a fan plenum and a plurality of upright supports having footpads at a lower end, the upright supports engaging the outwardly extending support beams; anda duct fluidly connecting the annular plenum of the lower portion with the fan plenum of the upper portion, the duct being disposed externally of the upper portion and the lower portion, and further being releasably attached to at least one of the upper portion and the lower portion;wherein the separable upper portion can be lifted away from the lower portion when the duct is disconnected from one of the upper portion and the lower portion and set on a flat surface and supported thereon by the upright supports. 14. The integrated head assembly of claim 13, wherein the upper portion further comprises a missile shield. 15. The integrated head assembly of claim 14, wherein the lower portion includes a ring beam that is attachable to the nuclear reactor. 16. The integrated head assembly of claim 15, wherein the ring beam further comprises a plurality of saddle members that distribute the load on the nuclear reactor. 17. The integrated head assembly of claim 13, wherein the lower portion further comprises at least one vertical channel to the annular plenum for cooling air flow. 18. The integrated assembly of claim 15, wherein the lower portion includes an air outlet aperture to the annular plenum, and the fan plenum includes an air inlet aperture, and wherein the plurality of fans are adapted to draw air from inside the lower portion, through the at least one vertical channel, to the annular plenum, and through the duct into the fan plenum. 19. The integrated assembly of claim 18, further comprising a chiller disposed in the fan plenum. |
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summary | ||
abstract | The present invention relates to a particle therapy apparatus used for radiation therapy. More particularly, this invention relates to a gantry for delivering particle beams which comprises means to analyse the incoming beam. Means are integrated into the gantry to limit the momentum spread of the beam and/or the emittance of the beam. |
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052232093 | summary | The invention relates to a method and a system for pressure relief of the containment of a nuclear power plant, having a filter disposed inside the containment which contains a washing fluid. Pressure relief of a nuclear power plant containment may become necessary if, contrary to every expectation, the pressure threatens to exceed the design value intended for the containment. In order to reduce the pressure in the containment of a nuclear power plant, a method and an apparatus disclosed in German Published, Non-Prosecuted Application DE 38 06 872 A1 may be used, in which a wet scrubber is provided as part of a filter inside the containment. A mixture of gas and steam produced inside the containment is cleaned of toxic and/or radioactive ingredients by the scrubber before the mixture is released into the environment. A special embodiment of a wet scrubber is described in German Published, Non-Prosecuted Application DE 38 15 850 A1. In that configuration, mixing of the gas and steam mixture with the washing fluid takes place inside Venturi nozzles, thereby assuring very intimate mixing and therefore a very good washing action. Once the design pressure of the containment is reached, regardless of the cause of its creation, the gas-steam mixture inside the containment amounts to approximately 70% water steam, along with compressed air and other gases formed inside the containment, which also includes a not-insignificant quantity of pure hydrogen with a probability bordering on certainty. In previously proposed embodiments having a wet scrubber accommodated inside the containment, the wet scrubber is either insulated and unheated or not insulated and accordingly continuously heated. As a result, the temperature of the washing fluid, such as water, may be markedly lower than the temperature of the gas-steam mixture, so that at least when the system is started up, upon passage of the gas-steam mixture through the washing fluid, an undesirably pronounced condensation of the water steam can occur. Even such condensation of the water steam is technically controllable only at considerable expense. However, a further disadvantage of the condensation of the water steam in the washing fluid is that as a result the relative proportion of the pure hydrogen gas in the gas-steam mixture flowing out of the wet scrubber can become undesirably high. With a continuously heated wet scrubber, there is an additional disadvantage which is that a considerable proportion of the washing fluid evaporates continuously and must be replaced. It is accordingly an object of the invention to provide a method and a system for pressure relief of the containment of a nuclear power plant having a wet filter provided inside the containment, which overcome the hereinafore-mentioned disadvantages of the heretofore-known methods and devices of this general type, and in which at most a non-relevant proportion of the water steam contained in the filter material condenses during the startup process. Since water is preferably used as the washing fluid, the temperature of the washing fluid must be approximately equal to the temperature of the filter material. On the other hand, as little washing fluid as possible should evaporate during continuous operation. With the foregoing and other objects in view there is provided, in accordance with the invention, a method for pressure relief of a containment of a nuclear power plant, which comprises heating a washing fluid in a filter disposed inside a containment at a relatively high rated heating power through a thermal bridge, with a gas-steam mixture filling the containment, prior to initial operation or initiation of operation of the filter; and rendering the thermal bridge substantially or virtually ineffective or even broken down completely in an operating state of the filter, leaving the washing fluid with only a remaining continuous rated heating power not being relevant for filtration. In accordance with another mode of the invention, there is provided a method which comprises setting the continuous rated heating power at less than 0.1 times the rated heating power. In accordance with a further mode of the invention, there is provided a method which comprises raising the washing fluid to an operating temperature after at most eight hours and preferably after substantially two hours of heating at the rated heating power. In accordance with an added mode of the invention, there is provided a method which comprises setting an operating temperature in a range of substantially from 100.degree. to 150.degree. C., raising the operating temperature to as high as substantially 260.degree. C. upon pressure relief directly from a primary loop of a nuclear power plant or direct pressure relief of a circuit that carries pressurized water, and supplying water to the filter as the washing fluid. In accordance with an additional mode of the invention, there is provided a method which comprises supplying a quantity of heat with the continuous rated heating power being less than a quantity of heat removed or drawn from the washing fluid by evaporation, for setting an operating temperature of the filter lower than an entry temperature of the gas-steam mixture. The method according to the invention is very advantageous, because as a result of the avoidance of concentration in the washing fluid, safe startup without significant changes of volume in the filter material is assured, and moreover a compact structure for the filter and a very small consumption of washing fluid are made possible. With the objects of the invention in view, there is also provided, in a nuclear power plant having a containment, a system for pressure relief of the containment, comprising a filter disposed inside the containment, the filter having a container, at least part of the container being double-walled or having two walls defining a chamber between the walls, and a heat-conducting fluid at least partly filling the chamber during a heating period and being at least half evaporated after attainment of an operating temperature. In accordance with another feature of the invention, the container has double-walled portions with the two walls, the double-walled portions having a heat-conducting resistance when not filled with the heat-conducting fluid being at least ten times higher than when filled with the heat-conducting fluid. In accordance with a further feature of the invention, the two mutually concentric walls define a double-walled, preferably cylindrical, middle part of the container surrounding or enclosing a vertical axis, and there is provided a single-walled curved base firmly closing the middle part toward the bottom, and a curved single-walled cap firmly closing the middle part toward the top. In accordance with an added feature of the invention, the double-walled middle part has a given height, the container includes a first inner chamber, the chamber between the walls is a second chamber having an annular cross section extending over the entire given height, and the second chamber communicates with the first inner chamber through openings formed just below the cap. This structure of the filter container makes it possible to fill the inner chamber with washing fluid to an extent that fluctuates within wide limits, and to fill the second chamber at least in its lower part with heat-conducting fluid, which represents part of a thermal bridge between the surroundings of the filter container and the filter contents, and evaporation of the heat-conducting fluid inside the second chamber is possible. Due to the disappearance of the heat-conducting fluid from the second chamber, the thermal bridge between the surroundings of the filter container and its contents is interrupted, so that the above-mentioned advantages are attained. After the disappearance of the heat-conducting fluid from the second chamber, this chamber is available to receive radioactive waste filtered out of the gas-steam mixture, so that this waste advantageously remains inside the containment. In accordance with an additional feature of the invention, there is provided washing fluid filling substantially 30 to 80% and preferably approximately 50% of the first inner chamber, the second chamber having a lower part being filled with the heat-conducting fluid and an upper part, and filter mats filling at least the upper part. In accordance with again another feature of the invention, the two walls have portions surrounding the heat-conducting fluid, and at least part of the second chamber has heat transfer fins increasing surface area in the vicinity of the wall portions. In accordance with again a further feature of the invention, both the washing fluid and the heating-conducting fluid are water. In accordance with again an added feature of the invention, there is provided a convection barrier in the form of a horizontal partition disposed in the second chamber above the heat-conducting fluid, the horizontal partition being perforated and/or slit. In accordance with again an additional feature of the invention, there are provided mist separators disposed in the upper part of the inner chamber in front of the openings leading to the second chamber. In accordance with still another feature of the invention, there is provided a vertical pipe disposed centrally in the first inner chamber for delivering a gas-steam mixture to be filtered, the vertical pipe having upper and lower ends, a horizontally extending segment through which the upper end of the vertical pipe communicates with the interior of the containment, radially disposed horizontal feed pipes into which the lower end of the vertical pipe discharges, and short Venturi nozzles communicating with the horizontal feed pipes just above the base. In accordance with still a further feature of the invention, the two walls are in the form of an inner wall and an outer wall, the outer wall has another opening formed therein leading to the outside, and there is provided a pipe penetrating the containment and communicating with the other opening for carrying a filtered gas-steam mixture out of the second chamber through the outer wall at approximately half the height of the filter mats. In accordance with a concomitant feature of the invention, there is provided an overpressure line through which the interior of the filter communicates with a primary loop of the nuclear power plant or with the interior of the containment, the overpressure line having an overpressure valve being closed during normal operation. 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 a system for pressure relief of the containment of a nuclear power plant, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
039789578 | abstract | A hollow tubular guiding mast made up of interconnected axial sections and adapted for reciprocably guiding the gripper or charging body for a core reactor. The mast includes reinforcing spars extending axially therein in the form of interconnected flat plates which diverge in the outward direction of the mast. Guide means, such as rollers, are adjustably mounted on the plates and engage angularly related regions of the body to be guided in the mast. |
claims | 1. A collimator for use in an X-ray CT apparatus, the collimator comprising:a collimator module including a first scattered ray eliminating part and a second scattered ray eliminating part; andresin provided between the first scattered ray eliminating part and the second scattered ray eliminating part so as to fill any space between the first scattered ray eliminating part and the second scattered ray eliminating part and configured to hold the first scattered ray eliminating part and the second scattered ray eliminating part. 2. The collimator according to claim 1, wherein the resin has a hollow structure. 3. The collimator according to claim 2, wherein, the resin further has a reinforcing part provided in a region of the hollow structure and configured to reinforce the resin. 4. The collimator according to claim 1, wherein a plurality of the collimator modules are arranged at least in a one-dimensional direction. 5. The collimator according to claim 1, whereinthe resin has a first slit and a second slit formed therein and each shaped like a groove, andthe first scattered ray eliminating part and the second scattered ray eliminating part are formed by pouring a filler containing metal particles into each of the first and the second slits and subsequently hardening the filler. 6. The collimator according to claim 5, wherein the first slit and the second slit are formed by using a 3D printer. 7. The collimator according to claim 5, wherein the first slit and the second slit are formed along an X-ray incident direction. 8. The collimator according to claim 5, wherein a depth varies among the grooves of the first slit and the second slit. 9. The collimator according to claim 5, wherein a width varies among the grooves of the first slit and the second slit. 10. A collimator module that is arranged at least in a one-dimensional direction and that structures a collimator for use in an X-ray CT apparatus, the collimator module comprising:a first scattered ray eliminating part;a second scattered ray eliminating part; andresin provided between the first scattered ray eliminating part and the second scattered ray eliminating part so as to fill any space between the first scattered ray eliminating part and the second scattered ray eliminating part and configured to hold the first scattered ray eliminating part and the second scattered ray eliminating part. 11. A collimator for use in an X-ray CT apparatus, comprising:a collimator module including a first scattered ray eliminating part and a second scattered ray eliminating part; andresin provided between the first scattered ray eliminating part and the second scattered ray eliminating part and configured to hold the first scattered ray eliminating part and the second scattered ray eliminating part,wherein the resin has a hollow structure. 12. A collimator for use in an X-ray CT apparatus, comprising:a collimator module including a first scattered ray eliminating part and a second scattered ray eliminating part; andresin provided between the first scattered ray eliminating part and the second scattered ray eliminating part and configured to hold the first scattered ray eliminating part and the second scattered ray eliminating part, whereinthe resin has a first slit and a second slit formed therein and each shaped like a groove, andthe first scattered ray eliminating part and the second scattered ray eliminating part are formed by pouring a filler containing metal particles into each of the first and the second slits and subsequently hardening the filler. 13. A collimator module that is arranged at least in a one-dimensional direction and that structures a collimator for use in an X-ray CT apparatus, the collimator module comprising:a first scattered ray eliminating part;a second scattered ray eliminating part; andresin provided between the first scattered ray eliminating part and the second scattered ray eliminating part and configured to hold the first scattered ray eliminating part and the second scattered ray eliminating part,wherein the resin has a hollow structure. 14. A collimator module that is arranged at least in a one-dimensional direction and that structures a collimator for use in an X-ray CT apparatus, the collimator module comprising:a first scattered ray eliminating part;a second scattered ray eliminating part; andresin provided between the first scattered ray eliminating part and the second scattered ray eliminating part and configured to hold the first scattered ray eliminating part and the second scattered ray eliminating part, whereinthe resin has a first slit and a second slit formed therein and each shaped like a groove, andthe first scattered ray eliminating part and the second scattered ray eliminating part are formed by pouring a filler containing metal particles into each of the first and the second slits and subsequently hardening the filler. |
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summary | ||
description | This application claims the benefit of the filing date of EP Application Serial No. 08 166 450 filed on Oct. 13, 2008, the disclosure of this application is hereby incorporated herein by reference. The present invention relates to the use of a transponder in a commercial installation, particularly a power plant. In addition, the present invention relates to a method for performing maintenance, servicing and/or repair work on a component and/or on a switching or measuring point associated with said component in a commercial installation, particularly a power plant. Furthermore, the present invention relates to a commercial installation, particularly a power plant, which is equipped with a transponder that is used therein. Transponders are radio communication appliances which are increasingly often used for identifying a wide variety of objects. In comparison with conventional bar code patterns, which are likewise used for identification, transponders have a greater storage capacity. Transponders are designed such that data can be transferred to and read from the transponder by radio. Transponders are now available from various manufacturers and can be used in a mixed mode with other appliances, since they are compatible on the basis of their standardization. During its manufacture, each individual transponder is usually provided with an explicit, invariable and readable identifier in the form of a transponder identification information unit, which means that no two transponders exist with the same transponder identification information unit. By attaching a transponder, it is therefore possible to explicitly identify any objects. Information stored on the transponder is stored invisibly to humans, however, and therefore cannot be read by the human eye. On the basis of this, the use of transponders in power plants and many commercial installations in order to assist and simplify a wide variety of processes in maintenance work and maintenance measures is regarded with scepticism. A large amount of the maintenance work and maintenance measures in commercial installations is accompanied by a process of what is known as disconnecting a component before the start of the work and measures and reconnecting the component when the work is concluded. Disconnection is generally understood to mean producing a state in which no voltage is present on electrical installations or components, and also comprises the possibly manual performance of switching actions on mechanical installations, such as the operation of manually operated valves. Particularly before the start of work for maintenance work and maintenance measures, it is necessary for the component to be disconnected without error—that is to say for the component to be effectively “switched off”—so that the safe performance of the work in situ, i.e. on the component, is assured. When the component is reconnected, it is necessary to ensure that work performed in parallel on this component has also been concluded altogether before the component is connected and hence restarted. This reconnection is also referred to as normalization of the component. The process of disconnection has been assisted for a relatively long time by a computer-aided method which manages the planning and performance of disconnection operations on individual components and also entire procedural systems by creating disconnection lists with individual disconnection steps which are output on paper. Example activities for performing an individual disconnection step are the manual closure of a mechanical valve or the stopping of an electric motor with subsequent removal of the fuse in order to prevent inadvertent reconnection. The disconnection steps are performed using the disconnection plan in the commercial installation, are rendered visually recognizable in situ by attaching adhesive labels, signs, markers or the like and are confirmed as performed on the disconnection list by means of a signature. When all the disconnection steps associated with a component have been performed and this has been checked using the disconnection list, it is possible to give the go-ahead for performing maintenance work and maintenance measures on the relevant component while ensuring a safe work area. The maintenance work and maintenance measures themselves should likewise be performed and confirmed using a list. Only after checking whether all the work and measures to be performed have been concluded can the relevant component be reconnected. The component is reconnected (what is known as normalization) using the disconnection list which needs to be worked through. This also includes removing the attached adhesive labels, signs, markers or the like. This concludes the disconnection operation. In commercial installations, however, it is not uncommon for several disconnection steps to be simultaneously active on one component of the installation and furthermore for disconnection steps in several active disconnection plans to relate to the same technical component. In this case, it is necessary to ensure that this individual component is disconnected for as long as just one of the simultaneously performed pieces of work is still in progress. Before the normalization is performed, it is therefore necessary to check whether one or more components quoted in the disconnection steps of one disconnection are also included in other active disconnection operations. Should this be the case then it is necessary to exclude this step from the normalization. This task is difficult and potentially susceptible to error, since each disconnection step needs to be checked against each disconnection step in all other active disconnection plans. Although the use of a computer-aided method of conflict recognition as addressed above increases the reliability of the check significantly, computer-aided disconnection does not stretch to the actual location of the disconnection, which means that human error can result in the disconnection of a component despite precise guidelines. Furthermore, it is possible for the disconnection to be reversed for an incorrect component because an annotation or marker indicating the disconnection is not present on the component, is illegible or is inaccurate. Another source of error which can result in the disconnection of a component being reversed is harboured by paper-based evaluation of a disconnection plan as a result of the overlooking of still active disconnection states, so that an individual component is incorrectly normalized, i.e. connected or switched on. Another risk of unwanted reversal of the disconnection results from a marker which has been attached to the component or to a switching point for this component in this regard having fallen off or being unrecognizable or having been inadvertently removed. The present invention relates to providing a solution which allows the aforementioned drawbacks and risks to be avoided and an increase in occupational safety in a commercial installation, particularly a power plant, to be achieved. In particular, the present invention relates to the use of a transponder, which has a reception element, a transmission element, a visual display unit and a memory element and also a control element which is operatively connected to these, in a commercial installation, particularly a power plant, in an arrangement or positioning on or in proximity to a component or at a switching or measuring point associated with said component for the purpose of visually displaying an operating state for the component or the switching or measuring point associated therewith using the visual display of the transponder. In addition, the present invention relates to a method of the type denoted in more detail at the outset by virtue of a transponder, which has a reception element, a transmission element, a visual display unit and a memory element and also a control element which is operatively connected to these, being arranged on or in proximity to the component or at the switching or measuring point associated with said component, and the visual display unit of the transponder being used to visually display an operating state for the component or the switching or measuring point associated therewith on said visual display unit. Furthermore, the present invention relates to a commercial installation, particularly a power plant, by virtue of it being equipped with a transponder which is used according to The present invention allows the work and measures required for repairing and/or maintaining a component, for example, to be accompanied in computer-aided fashion up to the relevant component. The information stored on the transponder, which information is variable, is furthermore rendered visually ascertainable, visible and readable to the human eye on the visual display unit. The visual display unit therefore dispenses with the need for attachment of adhesive labels, signs, markers or the like, which eliminates the involvement for attachment and removal and reduces time and costs. In addition, the invention avoids the risk of any erroneous measure during repair work and maintenance measures on a component and hence increases occupational safety. The use of the invention, such as for the process of disconnection, therefore provides not only technical/economic advantages but also an increase in the reliability during implementation and a substantial gain in occupational and production safety on account of explicit identification of a component, storage of component-specific information in the form of state/operational information units and visually ascertainable display of information contained in the state/operational information units directly on a relevant component. Particularly advantageously, a transponder can be used when it is incorporated in a communication system and/or an information system. In one refinement, the invention is therefore distinguished by the use of a transponder for producing a communication system which comprises a read/write unit. A further refinement of the invention is characterized by the use of a transponder for producing an information system which comprises an electronic data processing (EDP) system. Such a communication system and/or information system which comprises a transponder is appropriate and expedient particularly when it is employed and used for maintenance, servicing and/or repair work on the component and/or the commercial installation, for which reason the invention continues to provide for the use of a transponder in this connection. In order to permanently associate the visual display unit with a component of a commercial installation, it is also expedient, in line with one refinement of the invention, if there is a fixed arrangement of the transponder on the component. The visual display of an operating state of the component or of the switching or measuring point associate with said component by the display unit of the transponder, as provided for by the invention, can be employed particularly in connection with installation, servicing and/or repair work on the component and/or the commercial installation. To implement such use, one advantageous refinement of the invention provides for the memory unit to permanently store a transponder identification information unit and to be used to store at least one state/operational information unit which relates to a component associated with the transponder, wherein the transponder identification information unit and the at least one state/operational information unit are read from the memory element by means of the transmission element, and the at least one state/operational information unit is supplied to the memory element by means of the reception element, and wherein the transponder is designed such that at least one portion of the information contained in the state/operational informational units is displayed on the visual display unit. In order to provide a person working on a component of an installation with the most specific and quickly ascertainable information possible regarding the operating state of the component, one refinement of the invention provides for the visual display unit to visually display or show only the number of state/operational information units, but not the state/operational information unit itself. As a result, during the process of disconnection, for example, it is possible to explicitly ascertain that when there are two active disconnections displayed on the visual display unit there must be no normalization, i.e. no reconnection, of the component, even if the disconnection step provides for this. Expediently, one advantageous development of the invention provides for a warning signal to be shown on the visual display unit by means of the control element when the number of state/operational information units is greater than zero. This means that in the case of active disconnection a visually ascertainable warning message is output which may be a graphic, for example, in order to draw attention clearly to the disconnection. In respect of use in commercial installations, particularly power plants, it is of particular advantage if the state/operational information units which can be stored in the memory element each comprise a piece of information regarding a connection or disconnection state of the component and/or recently performed servicing on the component and/or a recently performed official examination of the component and/or recently performed measured value recording for the component, which is likewise provided by the invention. In one particularly preferred embodiment of the present invention, the visual display unit is a display based on electronic ink with bistable display elements. These are chemically microcapsules which contain two different colour components of different charge which orient themselves in the electrical field. On the basis of the particle sizes and the viscosity of the system, there is no immediate relaxation back to a disorderly initial state when the electrical field has been switched off, and hence no loss of the information written on the display unit, but rather there may merely be a decrease in contrast. This ensures that information is displayed on the display unit, even when the transponder is in a zero-voltage state. Furthermore, no power source is required in order to operate the visual display unit. As examples of electronic ink, reference may be made to the products from the companies Gyricon and E-Ink Cooperation. These products of electrophoretic display units are microcapsules which contain charged dye particles. The individual particles are colourless, however, and smaller than in the case of other electronic inks. These electrophoretic display units are flexible, insensitive to shock and stable under pressure. In addition, the comparatively low actuation voltage limits the circuit complexity for the supply of power. Alternatively, other display technologies can also be used for the display unit, however, reference being made in this case to the known technologies FLCD (Ferroeletric Liquid Crystal Display), EASL (Electrically Addressable Smetic Liquid Display), ZBD (Zenithal Bistable Devices), CHLCD (Cholesteric Liquid Crystal Display) and OLED (Organic Light Emitting Diode), which are described in more detail in the document WO 2006/012997. The information displayed on the display unit may comprise at least one alphanumeric character and/or at least one graphic. By way of example, the display of alphanumeric characters allows specific information to be read off, whereas a graphic can be used to increase attention. The transponder may be a passive transponder which draws its requisite power from a transmitted signal, which is sent by a read/write unit which is compatible with the transponder, for example. This allows the transponder to obtain and draw its power required for operation solely from such a read/write unit. In comparison with an active transponder, this has the advantage that no power source, such as a battery, is required for the operation of the transponder. In particular, in line with a further refinement of the invention, the transponder is an RFID transponder. In this context, the abbreviation RFID stands for Radio Frequency Identification and means identification using electromagnetic waves. By way of example, the transponder may be an RFID transponder which has an operating frequency of 13.56 MHz, this allowing contactless identification and localization of the transponder using electromagnetic waves. As a result, it is also possible for data to be read in and out on the transponder contactlessly, the transmission and reception source needing to be brought to within approximately 3 cm, preferably 2 cm to 6 cm, of the transponder. The short distance is therefore of particular importance, since, by way of example, a plurality of transponders associated with corresponding components in a commercial installation may be arranged close together on a control panel and only a defined interval between the transmission and reception sources and the transponder allows explicit identification of the transponder. In one refinement of the inventive use of the transponder in a communication system, the read/write unit stores at least one state/operational information unit and at least one identification information unit, wherein each identification information unit has at least one associated state/operational information unit, the read/write unit can read a transponder identification information unit from the memory element of the transponder, and the read/write unit takes the read transponder identification information unit as a basis for transmitting at least one state/operational information unit from the read/write unit to the transponder. In this case, it is then also expedient if the content of the at least one state/operational information unit which is stored in the read/write unit can be altered by a user of the read/write unit before the transmission to the transponder. This is expedient particularly in the case of reconnection, since the user first of all normalizes the component, i.e. reconnects it, and then confirms this on the read/write unit, so that the status of the relevant state/operational information unit can be updated. So that the user of the read/write unit can see that he is actually performing the measure on the correct component, the invention also provides for the read/write unit to be designed to compare the transponder identification information unit read from the transponder with the at least one identification information unit stored in the read/write unit. In a further refinement of the use of the transponder in connection with the communication system, the invention provides that if the content of the transponder identification information unit and of an identification information unit matches then the read/write unit reads state/operational information units stored in the transponder and/or transmits at least one state/operational information unit which is associated with the identification information unit and which is stored in the read/write unit from the read/write unit to the transponder. As a result, during the process of disconnection, for example, the read/write unit can be used to establish whether further disconnection steps which are stored on the transponder in the form of the state/operational information units are active for the relevant component, the disconnection step to be performed by the user also being able to be added on the transponder to the list of disconnection steps stored thereon in the form of a state/operational information unit. Following the normalization or reconnection of an initially disconnected component, it is necessary for the relevant disconnection step stored on the transponder to be removed from the memory element of the transponder. To this end, one refinement of the invention provides that if the content of the transponder identification information unit and of an identification information unit matches then the read/write unit sends a control signal to the transponder, which control signal prompts the control element to erase a state/operational information unit stored in the memory element When there are a plurality of disconnection steps stored on the transponder, provision may be made for the read/write unit to output a warning message when the number of state/operational information units read from the transponder is greater than zero. This refers the user of the read/write unit specifically to a possible conflict between disconnection steps or disconnection states. The read/write unit may be a mobile portable computer in the form of a handheld PDA (Personal Digital Assistant) or a laptop. Alternatively, the read/write unit may also be in the form of a mobile telephone which has the same functionality as a PDA or laptop. This has the advantage that the user can conveniently carry the read/write unit with him and use it at different locations in the commercial installation. The memory power of today's mobile computers allows an extensive collection of information for the widest variety of components and work which needs to be performed. The use of such a mobile computer significantly increases convenience for the user and occupational safety. In this case, the integration of a reading device for transponders into a mobile computer of this kind is widely known today and therefore easy and inexpensive to implement. In one refinement of the inventive use of the transponder in an information system, at least one of the state/operational information units stored in a read/write unit is read from the read/write unit by the central EDP system, and state/operational information units to be stored in the read/write unit are transmitted from the central EDP system to the read/write unit. In this context, it is then also expedient for the central EDP system to be designed to process state/operational information units read by the read/write unit, particularly to compare them with state/operational information units stored in the EDP system. In this case, both the state/operational information units stored in the EDP system and the state/operational information units stored in the read/write unit are subjected to data alignment or to a bidirectional synchronization operation. This ensures that the EDP system always has the most up-to-date information about the components in the installation, so that the EDP system can produce disconnection lists either in preparation for planned measures or else ad hoc in the case of a fault in the installation. In this context, provision may be made for the visual display unit to display only the number of state/operational information units but not the state/operational information unit itself. This provides a person with the most specific and quickly ascertainable information possible regarding the operating state of a component. As a result, during the process of disconnection, for example, it is possible to explicitly ascertain that when there are two active disconnections displayed on the visual display unit, the component must not be normalized, i.e. reconnected, even if the disconnection step provides for this. In addition, the person is compelled to use a read/write unit if he needs to access all the information stored in the transponder. When the method is implemented, the control element can be used to show a warning signal on the visual display unit if the number of state/operational information units is greater than zero. As a result, a visually ascertainable warning message, which may be a graphic, for example, is output in the case of active disconnection, for example, in order to draw attention clearly to the disconnection. Provision may also be made for a read/write unit to transmit the state/operational information unit which is received by means of the reception element and which is stored in the memory element in the form of a signal and for the read/write unit to be brought into the frequency-dependent reception/transmission range of the transponder, preferably to within to 3 to 6 cm of the transponder, in order to transmit the signal. This allows explicit identification of the transponder and contactless reading-in and reading-out of data stored on the transponder. In addition, provision is expediently made for the transponder identification information unit permanently stored in the transponder to be read therefrom by the read/write unit before the signal is transmitted and for the transponder identification information unit to be compared with an identification information unit stored in the read/write unit. This ensures that the state/operational information unit to be transmitted to the transponder is also transmitted to the correct transponder. In this case, if the transponder identification information unit and the identification information unit match, the read/write unit can read state/operational information units stored in the transponder, and a warning message can be output by the read/write unit if the number of read state/operational information units is greater than zero. In the case of active disconnection, for example, this displays a perceptible and possibly visually ascertainable warning message to the user in order to increase the user's attention to the detected conflict. If the transponder identification information unit and the identification information unit match, a state/operational information unit which is stored in the read/write unit and which is associated with the identification information unit can be subjected to a processing step in which the status of this state/operational information unit is changed before a signal containing the changed state/operational information unit is transmitted to the transponder. This is necessary particularly in the case of reconnection, since the user first of all normalizes the component, i.e. reconnects it, and then confirms this on the read/write unit, so that the status of the relevant state/operational information unit can be updated. Following the normalization, i.e. connection, of a component, it is necessary for the relevant disconnection step stored on the transponder to be removed from the memory element of the transponder. To this end, if the transponder identification information unit and the identification information unit match, the read/write unit sends a control signal to the transponder, as a result of which a state/operational information unit stored in the memory element is erased by the control unit. Expediently, the storage of the transmitted state/operational information unit is followed by the visual display visually displaying only the current number of state/operational information units stored in the transponder but not the state/operational information unit itself. During the process of disconnection, for example, this means that it is possible to explicitly indicate that when there are two active disconnections displayed on the visual display unit there must be no normalization, i.e. no reconnection, of the component, even if the disconnection step currently being handled provides for this. So that it is actually possible for state/operational information units and hence information to be transmitted to the transponder, at least one state/operational information unit is transmitted to the read/write unit by a central EDP system before the signal is transmitted to the transponder. Following the transmission of the signal to the transponder, the changed state/operational information unit which is stored in the read/write unit can be read by the central EDP system. This implements central data and information management via the operating state of the individual components. Finally, as already illustrated above at least in part, one refinement of the method also characterizes the invention in that use of the transponder is implemented in accordance with one of Claims 2 to 15. The features which are cited above and which will be explained below can be used not only in the respectively indicated combination but also in other combinations. The framework of the invention is defined only by the claims. FIG. 1 schematically shows an information system, denoted overall by reference numeral 1, which comprises a transponder 2, a mobile read/write unit 3 and an EDP system 4. The proportions shown in the illustration between the transponder 2, the mobile read/write unit 3 and the EDP system 4 do not necessarily correspond to the actual relationships. The transponder 2 has an opening 5 which is in the form of a hole and through which it is possible to put a mounting element in order to allow the transponder 2 to be permanently fitted to a component of a commercial installation. Besides mechanical mounting options provided by the opening 5, the transponder 2 can also have a lateral face, which is remote from a visual display unit 6 of the transponder 2, fitted on and/or to the component of the installation by means of an adhesive or the like. In particular, the transponder is intended to be used in connection with a commercial power plant. The transponder 2, which is in the form of thin small plate (the small plate may be in rigid or supple form), also has a reception element 7, a transmission element 8 and a control element 9. The control element 9 is connected to the reception element 7, to the transmission element 8 and to the visual display unit 6 and comprises a memory element 10. The memory element 10 permanently stores a unique identifier, produced during the manufacture of the transponder 2, in the form of a transponder identification information unit, which may be a number. Besides the transponder identification information unit, the memory element 10 can be used to store a plurality of state/operational information units which relate to a component associated with the transponder 2. In this case, each state/operational information unit comprises information associated with the relevant component, such as its component name (pump unit), the position thereof within the commercial installation, a short description in the form of an operating instruction, technical data and the like. This information is also referred to as master data for the component. Furthermore, the respective state/operational information unit may contain a piece of information about a measure to be performed on the component and/or comprise a piece of information regarding a connection or disconnection state of the component and/or recently performed servicing on the component and/or a recently performed official examination of the component and/or recently performed measured value recording for the component. The transponder identification information unit and the state/operational information units are read from the memory element 10 by an appliance which is compatible with the transponder 2 using the transmission element 8. In addition, state/operational information units are supplied to the memory element 10 via the reception element 7. The visual display unit 6 of the transponder 2 is a display based on electronic ink with bistable display elements, the information displayed thereon or the displayed information comprising alphanumeric characters and a graphic, as can be seen in FIG. 3. The visual display unit 6 therefore visually displays a portion of the information contained in the state/operational information units. Alternatively, the visual display unit 2 may be designed on the basis of a different technology from that of electronic ink. The transponder 2 illustrated in this embodiment also has a recess 11. A communication operation between the transponder 2 and a compatible appliance in the form of a read/write unit 3 starts when a read/write pen on this appliance is brought to within 3 cm of the transponder 2. In this case, the recess 11 serves merely as an aid to putting on the read/write pen, so that communication between the transponder 2 and the compatible appliance 3 is ensured. Alternatively, the recess can also be used in the case of capacitive coupling between the transponder 2 and the compatible appliance for the purpose of supplying electrical power from the electrical field of the compatible appliance by means of electrodes in the circuit of the transponder 2. The aforementioned elements of the transponder 2 are the basic elements for this operation and should not be understood as conclusive. Optionally, the transponder 2 may additionally have a microprocessor, which means that the transponder 2 is programmable and can evaluate and process received state/operational information units. The transponder 2 used in this exemplary embodiment is a passive transponder which draws its required power from a transmitted signal. Alternatively, an active transponder can also be used, however, but this requires a separate power source (e.g. a battery or a solar cell) for operation. For a further alternative, the transponder 2 may also be an RFID transponder which has an operating frequency of 13.56 MHz, which allows contactless identification and localization of the transponder using electromagnetic waves. A wide variety of read/write units 3 which are in the form of mobile portable computers in the form of handheld PDAs or laptops are suitable as an appliance which is compatible with the transponder 2. These appliances are generally compatible with a transponder 2, as described here, on account of their standardization. The read/write unit 3 in the form of a mobile computer has a display apparatus 12 for visually displaying information or for displaying the state/operational information units and is carried by a user when measures need to be performed on components of a commercial installation. A computer program which can be executed on the read/write unit 3 can be used to manage state/operational information units stored on the read/write unit 3 and the status of said state/operational information units can be altered by the user of the read/write unit 3. The transponder 2 and the read/write unit 3 interchange information as indicated by the arrows in FIG. 1, so that both form a communication system. For the purpose of information interchange with the transponder 2, at least one state/operational information unit and at least one identification information unit identifying and/or verifying the transponder 2 are stored in the read/write unit 3, each identification information unit having at least one associated state/operational information unit for a component. For the purpose of information interchange or communication with the transponder 2, the read/write unit 3 reads the transponder identification information unit from the memory element 10 of the transponder 2. If the transponder identification information unit and the identification information unit match, the read/write unit 3 transmits the state/operational information unit associated with the identification information unit to the transponder 2. The term “information/operational information unit” used above comprises information regarding an active or activatable connection or disconnection state of an associated component. The term “transponder identification information unit” comprises a transponder identification “feature”, preferably in the form of an alphanumeric or binary combination, which is stored on the transponder and which explicitly identifies the transponder. The term “identification information unit” comprises such a transponder identification “feature” stored on the read/write unit 3. Before transmission to the transponder 2, a user can alter the content of the state/operational information unit stored in the read/write unit 3, so that the transponder 2 receives an updated state/operational information unit. In this case, the update relates to a measure which is performed on the component and which, by way of example, may be disconnection or reconnection. In addition, the read/write unit 3 also reads the state/operational information units stored in the transponder if there is a match between the content of the transponder identification information unit and the identification information unit. In the case of a disconnection operation, it is possible for the number of state/operational information units read to be greater than zero, which means that there are multiple disconnections for the present component. In this case, the user of the read/write device 3 receives a warning message which draws his attention to this conflict and informs him that the component must not be reconnected under any circumstances. In addition, the read/write unit 3 in this case sends a control signal to the transponder 2 which prompts the control element 9 to erase the state/operational information unit, stored in the memory element 10, which is associated with the identification information unit. So that the read/write unit 3 actually holds state/operational information units which need to be transmitted to the transponder 2, appropriate state/operational information units are transmitted from the central EDP system 4 to the read/write unit 3 in advance of the measure which is to be performed on a component and the associated information interchange between the transponder 2 and the read/write unit 3. Within the central EDP system 4, the circuit diagram for the commercial installation is stored in a database which is managed by a computer program. In the database of the system, the disconnection location of the component is linked to the explicit identification information unit of the transponder 2. In the case of a process of disconnection, for example, the central EDP system 4 undertakes the planning of disconnection operations for individual components and also entire procedural systems by creating disconnection lists with individual disconnection steps. The disconnection steps correspond to the state/operational information units which are transmitted to the read/write unit 3 after the planning and creation of the disconnection list. When the list has been processed, the state/operational information units stored in the read/write unit are read and processed by the central EDP system 4, with the read state/operational information units being compared with state/operational information units which are stored in the EDP system 4. In this context, measures which are not performed are recognized by the EDP system 4 and are incorporated into an appropriate list for fresh performance and processed at a later time. The text below describes an example of the equipment of a component of a power plant in a commercial installation and also an exemplary disconnection and connection process, in which the transponder 2, the communication system comprising the transponder 2 and the read/write unit 3, and the information system 1 comprising the transponder 2, the read/write unit 3 and the central EDP system 4 are used. Such a disconnection and connection process is performed in power plants in the course of servicing, maintenance and/or repair work. The “disconnection” is used to switch off the components or units and render them “safe”, so that it is possible to work on them or on downstream or associated components without risk. Following the conclusion of the work, the components are connected or switched on again. This operation is called normalization. Prior to the use of the information system 1 in a commercial installation, such as a power plant, in particular, a preparatory measure involves all the components of the commercial installation which possibly need to be operated in the course of a disconnection being permanently fitted with a respective transponder 2 having a visual display 5. The mobile read/write unit 3 on which an EDP application for assisting the method of disconnection is performed is used to read a transponder identification information unit from the transponder 2 and to associate this explicit and invariable identification identifier with other information from this component, such as a component name (pump unit), a location name in the installation, a short description in the form of an operating instruction, technical data and the like, which are also called master data. This preparatory measure also involves initialization of the data management on the transponder 2, in which a list of state/operational information is erased. The state/operational information units respectively contain an identification information unit, which is a number, for example and a piece of information which indicates a connection or disconnection state of the component, for example. Furthermore, the state/operational information units can also contain the master data for the relevant component. The conclusion of initialization is shown readably by the transponder 2 by displaying the digits “00” on the visual display unit 5, as shown in FIG. 2. The digits “00” indicate to the personnel in the commercial installation that this component is connected, i.e. that there is no disconnection of the component. Following the conclusion of the preparatory measure and hence of the association of the respective transponder identification information units with the relevant components, the respective information units are transmitted from the mobile read/write unit 3 to the central EDP system 4 and are stored thereon. As the result of the preparatory measure, the central EDP system 4 creates an installation plan in which each component has an associated explicit identification information unit and/or an associated transponder identification information unit. A disconnection process which is to be performed for a component can be rough divided into three sections. In a first step, what is known as a disconnection plan is created, followed in a second step by the performance of the disconnection, which is then followed in a concluding third step by the reconnection or normalization. For the disconnection which is to be performed for a component of the commercial installation, in the exemplary embodiment of a power plant, the EDP system 4 is used to produce a disconnection plan and/or a disconnection list with one or more disconnection steps, each individual disconnection step having an associated explicit identification information unit and transponder identification information unit. Each individual disconnection step is produced as a state/operational information unit in the EDP system 4 and incorporated into the disconnection list. To assist the process of disconnection, the disconnection list with the state/operational information units which are produced in the EDP system 4 and which respectively contain the identification information unit, the master data for a component which are associated with the identification information unit and the measure which is to be performed on the component, is transmitted to at least one mobile read/write unit 3. In this case, a possible split in the disconnection list can be made according to different aspects. It is usual to separate mechanical and electrical disconnection, because these two measures require people with different qualifications and can usually also only be performed by these people. The display 12 of the mobile read/write unit 3 is used to display the disconnection list with all the disconnection steps which are to be performed which are associated with one person. Suitable means, for example a colour envelope or a symbol, are used to identify already performed disconnection steps on the display 12 of the read/write unit 3. Furthermore, the display 12 is used to display the master data for a respective component or can be used to show them by means of user interaction. An online connection between the mobile read/write unit 3 and the central EDP system 4 is not required, but can be set up optionally, for example by means of WLAN. Since reception cannot be safeguarded everywhere in commercial installations, however, autarkic operation (without an online connection) of the mobile read/write unit 3 is provided primarily. Upon reaching a component of the commercial installation which needs to be disconnected, the user holds the read/write unit 3 close to the transponder 2 which is attached to the component or in proximity thereto or else to a switching point therefor. As a result of comparison of the transponder identification information unit read from the transponder 2 with the identification information units stored in the disconnection list, which each have a corresponding associated disconnection step, the correct disconnection step is automatically shown or, in the event of an error, a message is displayed if this component is not associated with a disconnection step. When the user of the read/write unit 3 has in this way satisfied himself that he is at the correct component, he disconnects it. The component can be disconnected by pushing a switch or rotating a crank, for example. The performance of the disconnection measure is confirmed by the user by operating a button on the mobile read/write unit 3. Next, the read/write unit 3 is brought close to the transponder 2 of the component, so that the read/write unit 3 transmits a state/operational information unit (for example the identification number of the disconnection plan which is currently to be executed or initialized) to the transponder 2 and stores the state/operational information unit possibly in a convenient list of the disconnection plan numbers which are stored in the transponder 2. The transponder 2 stores this state/operational information unit permanently in its memory element 10. In response to the data transmission, the transponder 2 also uses the visual display unit 5 to show a conspicuous symbol 13 and displays the number of simultaneously active disconnection operations as a number. Accordingly, one active disconnection operation is displayed in FIG. 3, whereas in FIG. 4 two active switching operations are displayed. This information is therefore visually accessible to everyone and no read/write unit 3 is required for this. The disconnection plan numbers will usually not be shown on the display, which means that this list can still be read only using the read/write unit 3. Following the conclusion of the disconnection operations, the confirmations are transmitted from the mobile read/write unit 3 to the central EDP system 4. Complete processing of the disconnection list stored on the read/write unit 3 is desirable, but not necessary. The central EDP system 4 performs the confirmation of the disconnection steps from a plurality of mobile read/write units 3 and shows the combined status of the disconnection operations. Disconnection steps which have not yet taken place are therefore reliably recognized and can be reassigned to a mobile read/write unit 3. Only when the central EDP system 4 reports that all the disconnection steps of a disconnection plan have been performed is it possible to grant a work approval for the component. As soon as the work associated with a disconnection plan is reported to have ended, the disconnection plan can be selected in the EDP system 4 for normalization and can be transmitted in the same way as for the disconnection to one or more mobile read/write units 3. When the component to be normalized is reached, the read/write unit 3 is brought close to the transponder 2 which is attached to the component. During this operation, the mobile read/write unit 3 reads the transponder identification information unit from the memory element 10 of the transponder 2 and the list of disconnection plan numbers which are stored on the transponder 2 in the form of state/operational information units. By comparing the read transponder identification information unit with the identification information units stored for the disconnection steps, the correct disconnection step is sought and displayed using an EDP application installed on the read/write unit 3. In the event of an error, i.e. if this component is not associated with any disconnection step, a message is displayed on the display 12 of the read/write unit 3. In the next step, the application on the mobile read/write unit 3 checks whether the list of disconnection plan numbers contains exclusively the number of the disconnection plan which is to be normalized. Should this not be the case, a conspicuous message 13 is displayed on the display 12 which provides information that other disconnections are still active and normalization must not take place. When it has been ascertained that the correct component is involved and the normalization can also be performed, the component is put back into the normal state, i.e. it is connected by pushing a switch or rotating a crank, for example. The performance of the measure is confirmed by operating a button on the mobile read/write unit 3 and then holding the read/write unit 3 against the transponder 2 of the component. During this operation, the read/write unit 3 erases the relevant state/operational information unit which is stored in the memory element 10. The transponder 2 stores this information permanently in its memory element 10. In response to the data transmission, the transponder 2 updates the visual display unit 6. The conspicuous symbol 13 continues to be displayed if further disconnections are active, otherwise it disappears. The number of simultaneously active disconnection operations is updated as a number and as a result provides the user with a visual report. The current state of the disconnections therefor continues to be visually accessible to everyone. A read/write appliance 3 is not required for this purpose in order to identify the number of active disconnection states which is shown in FIGS. 3 and 4. Following the conclusion of the normalization operations, the confirmations are transmitted from the mobile read/write unit 3 to the central EDP system 4. In a similar manner to the practice with a disconnection, the EDP system 4 tracks the normalization operations performed and signals which normalization operations are still outstanding and/or for which disconnection operations the normalization has been completed. The invention can be used not only for a disconnection process for components of a commercial installation, such as a power plant, as described above, however, but can also be used for other visually displayable measures, which ensures an increase in the reliability of implementation and a substantial gain in occupational and production safety. In this context, the transponder 2 is in turn fitted to the component itself or at a location suitable for the measure. In FIG. 5, the visual display unit 6 of the transponder 2 displays recently performed servicing for a component of a commercial installation. In this case, the visual display 6 is used to show a symbol 14 which represents a servicing measure. The other information shown in the visual display 6 shows the date of the recently performed servicing measure. In equivalent fashion to the process of disconnection, the central EDP system 4 creates and processes a servicing plan. When it has been processed, said servicing plan is stored on the read/write unit 3, which involves the relevant state/operational information unit being changed by the user of the read/write unit 3. Following conclusion of the servicing work, this changed state/operational information unit is transmitted to the transponder 2, which then updates the date shown for the recently performed servicing. As further information, FIG. 5 shows the relevant servicing plan “WP 07” on the visual display unit 6, since the servicing of a component or machine in a commercial installation has several associated servicing plans. In the case shown, servicing according to servicing plan “WP 07” was accordingly recently performed on Feb. 6, 2008. In FIG. 6, the visual display unit 6 of the transponder 2 shows a recently performed official examination (Technical Control Board) for the component whereas FIG. 7 displays recently performed measured value recording for the component. A change to the information shown in FIG. 6 and FIG. 7 is made in identical fashion to the previously described measure of recently performed servicing, so that reference is made to the above explanations in this regard. |
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claims | 1. A method of manufacturing an antiscattering grid comprising:providing a substrate made of a polymer material which is formed by radiation curing a monomer sensitive to the radiation;forming the substrate into a grid by radiation curing the substrate exposed to the radiation to have a plurality of cured partitions which define a plurality of cells passing through the substrate; andproviding a layer of metal on a surface of the substrate, thereby resulting in the antiscattering grid comprising the plurality of cured partitions having metal-layered cells passing through the substrate;wherein the metal-layered cells define open-air openings passing completely through the antiscattering grid. 2. The method according to claim 1 wherein the metal layer provided on the surface of the substrate is formed by chemical vapor deposition. 3. The method according to claim 1 wherein:the metal layer is provided on all of the surfaces of the substrate; andthe metal on external surfaces of the substrate is removed in order to leave only the metal covering the partitions inside the cells. 4. The method according to claim 3 wherein the metal layer on the external surfaces of the substrate is removed by abrasion. 5. The method according to claim 3 wherein the metal layer on the external surfaces of the substrate is removed by plasma etching. 6. The method according to claim 3 wherein the metal layer on the external surfaces of the substrate is removed by laser ablation. 7. The method according to claim 1 wherein the providing a metal layer step comprises providing a metal layer which covers the partitions inside the cells but not on external surfaces of the substrate outside the cells. 8. The method according to claim 7 wherein the external surfaces of the substrate outside the cells are masked during the providing a layer of metal step so as to provide the metal only on the partitions inside the cells. 9. The method according to claim 1 wherein the metal comprises at least one layer formed by a metal selected from the group consisting of gold, copper, tantalum and lead. 10. The method according to claim 1 wherein the monomer is a resin selected from the group consisting of epoxy and acrylic. 11. The method according to claim 1 wherein the substrate has a substantially planar shape. 12. The method according to claim 1 wherein the cells have a polyhedral shape. 13. The method according to claim 1 wherein the partitions are formed to orient as a focused grid. 14. The method of claim 1, wherein the curing technique used is stereolithography. 15. The method according to claim 14 wherein the substrate is formed by point-by-point stereolithography. 16. The method according to claim 14 wherein the substrate is formed by whole-layer stereolithography. 17. The method of claim 1, wherein the providing a layer of metal comprises:depositing a layer of metal upon the substrate such that a cross section of the antiscattering grid comprises polymer material disposed between two instances of the layer of the metal deposited and the cells defining open-air openings are disposed between another two instances of the layer of metal deposited. 18. The method of claim 1, further comprising:subsequent to the providing a layer of metal, filling the cells with a polymer. 19. A method of manufacturing an antiscattering grid comprising:providing a substrate made of a polymer material which is formed by radiation curing a monomer sensitive to the radiation;forming the substrate into a grid by radiation curing the substrate exposed to the radiation to have a plurality of cured partitions which define a plurality of cells passing through the substrate;providing a layer of metal on a surface of the substrate; andfilling the cells with a polymer. 20. The method according to claim 19 comprising filling the cells with a polymer which is similar to the polymer of the substrate. 21. A method of manufacturing an antiscattering grid comprising:providing a substantially planar substrate made of a polymer material of a given thickness and sensitive to radiation;forming in the substrate by radiation curing of the polymer material a plurality of partitions which define a plurality of cells passing through the substrate and are oriented to a focal point; andproviding a layer of metal on a surface of the substrate, thereby resulting in the antiscattering grid comprising the plurality of partitions having metal-layered cells passing through the substrate;wherein the metal-layered cells define open-air openings passing completely through the antiscattering grid. 22. The method according to claim 21 wherein the ratio of the grid thickness to the distance between partitions is a ratio of greater than 8. 23. The method according to claim 21 wherein the cells are formed to have an opening of about 200 μm to about 300 μm and the partitions are formed to have a thickness of about 50 μm to about 100 μm. 24. The method according to claim 21 wherein the cells are formed to have a pitch of about 50 μm to about 100 μm and the partitions are formed to have a thickness of about 20 μm to about 50 μm. 25. The method according to claim 21 wherein the cells are formed to have a shape of a parallelogram. 26. The method according to claim 21 wherein the substrate is formed to have a thickness that is not constant. 27. The method according to claim 21 wherein the grid is formed to have a thickness of greater than about 1.6 mm and less than about 3 mm. 28. The method according to claim 21 wherein the grid is formed to have a thickness of greater than about 0.4 mm and less than about 1 mm. 29. The method according to claim 21 wherein the cells are formed to have a quasi-periodic pattern. 30. The method according to claim 21 wherein the cells are formed to have a pattern in which the pitch or the period varies continuously. 31. A method of manufacturing an antiscattering grid comprising:providing a monomer precursor fluid sensitive to and curable by radiation;curing the fluid by radiation to form a substrate defining a grid, wherein the fluid exposed to the radiation is cured to form a plurality of cured partitions which define a plurality of cells passing through the substrate; andproviding a layer of metal on a surface of the substrate thereby resulting in the antiscattering grid comprising the plurality of cured partitions having metal-layered cells passing through the substrate;wherein the metal-layered cells define open-air openings passing completely through the antiscattering grid. 32. The method of claim 31, wherein the providing comprises:providing the metal such that the metal covers only the partitions inside the cells. 33. The method of claim 31, wherein:the cells have a polyhedral shape. 34. The method of claim 31, wherein:the partitions are formed to orient as a focused grid. 35. The method of claim 31, wherein:the cells are formed to have a quasi-periodic pattern. 36. The method of claim 31, wherein:the cells are formed to have a pattern in which the pitch or the period varies continuously. 37. The method of claim 31, wherein:at least some of the partitions separating the cells are misaligned. 38. The method of claim 31, wherein:the substrate is a one-piece substrate. 39. The method of claim 31, wherein:the metal layer provided on the surface of the substrate is formed by chemical vapor deposition. 40. The method of claim 31, wherein the curing comprises stereolithography. 41. A method of manufacturing an antiscattering grid comprising:providing a substrate made of a polymer material which is formed by radiation curing a monomer sensitive to the radiation;forming a polymer lattice by radiation curing the substrate exposed to the radiation, the polymer lattice comprising a plurality of cured polymer partitions that define a plurality of interior walls of open-air through cells passing completely through the polymer lattice; andproviding a layer of metal on a surface of the polymer lattice, thereby resulting in the antiscattering grid comprising a metal-layered polymer lattice comprising the plurality of cured polymer partitions that define a plurality of metal-layered interior walls of open-air through cells passing completely through the antiscattering grid. 42. The method of claim 41, wherein the providing a layer of metal comprises:depositing a layer of metal upon the polymer lattice such that a cross section of the antiscattering grid comprises polymer material disposed between two instances of the plurality of metal-layered interior walls and air disposed between another two instances of the plurality of metal-layered interior walls. 43. The method of claim 41, further comprising:subsequent to the providing a layer of metal, filling the open-air through cells with a polymer. 44. The method according to claim 41 wherein the metal layer provided on the surface of the substrate is formed by chemical vapor deposition. 45. The method according to claim 41 wherein:the metal layer is provided on all of the surfaces of the polymer lattice; andthe metal on external surfaces of the polymer lattice outside of the cells is removed in order to leave only the metal covering the plurality of interior walls inside the cells. 46. The method according to claim 45 wherein the metal layer on the external surfaces of the polymer lattice is removed by abrasion. 47. The method according to claim 45 wherein the metal layer on the external surfaces of the polymer lattice is removed by plasma etching. 48. The method according to claim 45 wherein the metal layer on the external surfaces of the polymer lattice is removed by laser ablation. 49. The method according to claim 41 wherein the providing a metal layer step comprises providing a metal layer which covers the plurality of interior walls but not external surfaces of the polymer lattice outside of the cells. 50. The method according to claim 49 wherein the external surfaces of the polymer lattice outside of the cells are masked during the providing a layer of metal step so as to provide the metal only on the plurality of interior walls. 51. The method of claim 41, wherein the curing comprises stereolithography. 52. The method according to claim 51 wherein the substrate is formed by point-by-point stereolithography. 53. The method according to claim 51 wherein the substrate is formed by whole-layer stereolithography. |
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055464351 | summary | This invention relates to a method of carrying out leak detection of a fuel assembly for a pressurized-water reactor (PWR). BACKGROUND OF THE INVENTION There are substantially two modern method for carrying out leak detection of a fuel assembly for nuclear light-water reactors. INMAST-sipping for PWR-reactor plants and TELESCOPE-sipping for BWR-reactor plants (BWR=Boiling-Water Reactor). A leakage occurring leads to the reactor water and hence the different parts of the primary circuit of the reactor being contaminated with the radioactive fission products. When a contamination of the reactor water has been determined or is suspected to have occurred, it is of the utmost importance that the leakage is locted so that leaky fuel assemblys can be replaced and later repaired. A fuel assembly includes most often several fuel rods. The so called INMAST-sipping is used for PWR-reactor plants and is a sipping method of conventional type, where a so called on-line gas detection is used. At the INMAST-sipping, the fuel assembly suspected to be leaking is drawn into a suction hood, which preferably is the refuelling machine mast. The fission gases emitted from the fuel assembly are sampled in the upper part of the hood, and thereafter detected in a gas detection circuit. The gas volume in which an amount of fission gas from a damaged fuel assembly is emitted is quite voluminous. This gives the result that the detection sensitivity is quite low. The FR-2509898 describes a fuel leak detection for a PWR in which the unit of a fuel assembly drawn into a suction hood is lifted up several meters well above the reactor core in order to increase the internal relative pressure of the rods in relation to the water pressure but so that the fuel assembly is still surrounded by water. The gas emitted in the water and then in a gas volume above the water level in the hood is sucked from the upper part of the hood, which then is closed but for the gas sucking device. The possible content of radiactive fission products in the sucked gas is examined. The detection sensitivity is quite low. In order to increase the same a gas stream is forced through the water around the fuel assembly in the hood from its bottom to the top. However, the detection sensitivity is quite low even with this measure. At the so called TELESCOPE-sipping, which is totally adapted to a BWR-reactor plant, a great amount of the water surrounding a damaged fuel assembly is pumped from a nozzle placed at or a doom placed around the upper part of the fuel assembly to be examined lifted up somewhat from the reactor core by a gripper at a telescope mast arrangement from the reactor core and water pumped from the nozzle is supplied to a measuring circuit having a little volume. The water is degased in order to make a gas detection on-line. The detection has a high measuring sensitivity. This kind of sipping is described in the Swedish Patent No. 91015065, according to which a hood or a nozzle arrangement is placed in the region around the gripper at the upper part of a lifted fuel assembly. A pump sucks water from that region. The fuel assembly of a BWR-reactor plant is in itself closed to the surroundings so that the pumped water from the nozzle to a great extent comes from water streaming through the inside of the assembly. The fuel assembly lifted-up is through-flushed with reactor water and the gripper is rinsed-off. This is also the case when the fuel assembly, after having been lifted to a given position vertically, is held in this position or is relowered, and the analysis of leaking fission gases is then carried out. The fuel assemblies in a PWR-reactor plant have quite another open structure than the closed assemblies in a BWR-reactor plant. A nozzle at the upper part of a fuel assembly of a PWR could not possibly have the function to collect water streaming around the fuel rods in the fuel assembly. The radiation is emitted from its sides. Therefore, the fuel assemblies of a PWR-reactor plant must be placed in a closed space when lifted from the reactor core. An object of the invention is to provide a fuel leak detection for a PWR with a high sensitivity. According to the invention the gas measuring circuit at an INMAST-sipping device is replaced by a water sampling circuit. Water surrounding the fuel assembly to be examined and then placed inside a lifting rod is pumped from the water inside of the rod to a water/gas separation device. The separated gas is then detected on-line in the same way as for the so called TELESCOPE-sipping for BWR-reactor plants. |
051857758 | summary | BACKGROUND OF THE INVENTION 1.Field of the Invention The present invention relates to X-ray apparatuses, and more particularly, within such apparatuses, it relates to a device or filter for spatially modifying exposure to X-rays as a function of the morphology of the patient's body or of the portion of the body receiving X-rays. 2.Description of the Prior Art The invention is more particularly intended for X-ray apparatuses that are used for angiographical examination of the lower limbs. It is recalled that angiography is the application and adaptation of the radiological technique to the vascular network: arteries, veins, tissues under perfusion. It makes use of "contrast" liquids based on iodine which are opaque to X-rays and which are injected into the vascular network in order to enable it to be visualized by distinguishing it from the surrounding tissue. More precisely, the patient is laid on a table which is designed to move relative to a source of radiation associated with a receiver disposed on the other side of the patient from the source. The practitioner injects the X-ray opaque liquid, known as the "contrast" liquid, into an artery or a vein of the patient lying on the table. Then, a few seconds after the injection, a plurality of successive X-ray pictures of the patient are taken so as to visualize and measure the progress of the contrast substance along the blood vessels as a function of time. When angiographic examination is applied to the lower limbs, i.e. over a length of about 120 centimeters, several methods are currently in use. The first method consists in using an "arteriophlebograph" as the receiver equipment, which equipment comprises a hexagonal drum with six film and reinforcing screen pairs provided thereon, each pair being 120 cm long and 35 cm wide, thereby enabling six exposures and thus six photographic pictures to be obtained at different instants. In this method, each picture provides an image over the full length of the limbs and the patient therefore receives a considerable dose of X-rays since the entire lower portion of the body is exposed when each picture is taken. In addition, since the exposure is the same over the entire lower portion of the body, there is a large difference in contrast between the abdomen and the feet, given the differences in X-ray absorption presented by these portions of the body. This large difference in contrast makes it difficult to identify the contrast substance. In order to reduce this contrast difference, proposals have been made to place a rudimentary filter on the path of the X-ray beams, the filter being of the wedge type, i.e. being constituted by a sheet of varying thickness, thereby attenuating X-rays at the feet more than at the abdomen. Proposals have also been made to use a reinforcing screen whose X-ray to light photon efficiency varies over the length of the lower limbs, so as to present low efficiency at the feet and medium to high efficiency at the abdomen. The changes in contrast that are obtained using such devices, i.e. a wedge-shaped plate or a varying efficiency reinforcing screen, are fairly coarse. A second method consists in examining the lower limbs zone by zone, with each zone corresponding, for example, to an effective field of 35 cm.times.35 cm or to a diameter of 30 cm or of 35 cm. To do this, relative displacement is obtained between the X-ray beam and the patient between taking successive pictures such that each picture gives a partial image of the limbs, and the resulting set of images covers the entire length of the limbs. In this second method, the exposure parameters vary from one picture to the next in order to take into account of the variations in thickness between the abdomen and the feet. In practice, given the dynamic range of the variations required, it is necessary to vary both the supply voltage kV applied to the tube of the X-ray source and the product mA.s of the anode current mA multiplied by the exposure time s. This therefore results in a change in the contrast between the images taken of different zones of the patient, and this makes identifying the contrast substance difficult, and more generally makes analyzing the images difficult, particularly if they are processed digitally. In both methods outlined briefly above, some portions of the sensitive surface of the receiver, corresponding to the gap between the legs or to margins outside the legs are subjected to X-rays which are unattenuated and this degrades image quality. This problem is solved by placing various absorbent bodies such as plastic cylinders filled with water around the patient and between the patient's legs. Manipulating such cylinders is inconvenient both for the patient and for the practitioner. The object of the present invention is therefore to provide an X-ray apparatus which avoids the above-mentioned drawbacks. This result is achieved by using a filtering or absorption device which is disposed between the source of radiation and the receiver, and preferably in the vicinity of the source. SUMMARY OF THE INVENTION The present invention provides a radiological apparatus, in particular for radiological examination of the lower limbs, the apparatus comprising an X-ray source which emits a beam of X-rays towards a patient, a table on which the patient may be laid out, and an X-ray receiver disposed on the opposite side of the table to the side on which the source is placed, the table and/or the source-receiver pair being capable of displacement relative to each other so that the X-ray beam is capable at least of scanning the lower limbs of the patient in the longitudinal direction, said apparatus further including an X-ray attenuation filter device which is disposed between the X-ray source and the receiver, the said device providing attenuation on each X-ray path in the beam such that the total attenuation to which the X-rays are subjected on any of the paths to the receiver is substantially the same for all of the paths in the beam, thereby homogenizing the exposure of the image-forming receiver. Preferably, the filter is disposed close to the source of radiation and first means are provided to displace the filter perpendicularly relative to the beam in such a manner that the said beam intersects corresponding portions of the filter and of the patient's body. According to another feature of the invention, second means are provided to modify the ratio between the source-filter distance and/or the filter-patient distance in such a manner as to adapt the size of the filter to the size of the patient. According to another feature of the invention, third means are provided for displacing the filter horizontally over a distance which is different from the advance step size defined by the scale factor relating the size of the patient to the length of the filter. According to another feature of the invention, provision is made for using a plurality of interchangeable filters, with each filter being adapted to the morphology of the patient to be examined. |
description | Generally, an electronic article such as a system on chip (SOC) is associated with an interface that enables the SOC to electrically connect with a part, such as a dynamic random access memory (DRAM). Routinely, however, there are multiple vendors available to provide the part, where the part varies somewhat from vendor to vendor such that the manner or mechanism for electrically connecting the part to the SOC varies depending upon the vendor chosen to supply the part. Accordingly, different interfaces are generally required to electrically connect the part to the SOC when different vendors are used. That is, a first interface is needed to electrically connect the part to the SOC when the part is obtained from a first vendor and a second interface is need to electrically connect the part to the SOC where the part is obtained from a second vendor. Generally, an interface comprises one or more sets of features. A mask of a set of masks is typically used to form, on the SOC, a set of features of the sets of features. For example, where vendor A is selected to provide DRAM A, a first set of masks ‘A’ comprising a first mask A, a second mask A, and a third mask A is used to form an interface A compatible with the DRAM A. The first mask A is used to form a first set of features A within the interface A, the second mask A is used to form a second set of features A within the interface A, and the third mask A is used to form a third set of features A within the interface A. When a switch is made from DRAM A to DRAM B provided by vendor B, a second set of masks ‘B’ comprising a first mask B, a second mask B, and a third mask B is required to fabricate interface B on the SOC such that the SOC is compatible with the DRAM B. The first mask B is used to form a first set of features B within the interface B, the second mask B is used to form a second set of features B within the interface B, and the third mask B is used to form a third set of features B within the interface B. It will be appreciated that the first set of masks ‘A’ is specific to vendor A and cannot be used to fabricate interface B. Similarly, the second set of masks ‘B’ is specific to vendor B and cannot be used to fabricate interface A. This summary is provided to introduce a selection of concepts in a simplified form that are further described below in the detailed description. This summary is not intended to be an extensive overview of the claimed subject matter, identify key factors or essential features of the claimed subject matter, nor is it intended to be used to limit the scope of the claimed subject matter. One or more embodiments of techniques or systems for incorporating a common template into an electronic article design, such as a system on chip (SOC) design, are provided herein. It will be appreciated that while SOC and the like are substantially referred to herein, the instant application is not to be so limited. That is, the instant application, including the scope of the appended claims, is not necessarily limited to a SOC, a SOC design, etc. Rather, more than merely a SOC, a SOC design, etc. are within the contemplated scope of the present disclosure. Generally, a vendor specific interface is an interface that is formed on a SOC and that enables the SOC to mate or electrically connect with a part, such as a DRAM, from a particular vendor, where the vendor specific interface would not allow the SOC to electrically connect to the part if the part were obtained from a different vendor. The vendor specific interface is, for example, fabricated using a set of one or more masks and comprises one or more sets of corresponding features. According to some aspects provided herein, the SOC is designed such that at least a portion of the vendor specific interface fabricated on the SOC is standardized across one or more vendors. Because at least a portion of the vendor specific interface fabricated on the SOC is standardized across one or more vendors, the vendor specific interface is at time merely referred to as an interface, as opposed to a vendor specific interface. The standardized portion of the interface is the same regardless of the vendor selected to provide the part, or rather is the same for at least two vendors. The standardized portion of the interface is standardized because a same set of masks is used to fabricate the standardized portion on the SOC. In some embodiments, the standardized portion of the interface is regarded as a common template. Similarly, the same set of masks used to fabricate the standardized portion on the SOC is regarded as a common template mask set. The common template mask set comprises one or more common template masks, where a common template mask is used to form a set of features of the common template. The common template meets design requirements for multiple vendors by comprising features, corresponding to polygon positions, for example, that allow the SOC to be electrically connected to the part regardless of the vendor that provided the part. It will be appreciated, however, that in some instances a vendor specific layer or set of features is required depending upon the vendor that is providing the part. The common template mask set similarly meets design requirements for multiple vendors by producing the features of the common template. Given the common template mask set, or parameters thereof, a set of design rules is generated based on the common template mask set. The set of design rules is provided to a third party, such as a customer or any type of consuming entity interested in being able to connect a SOC to a part, such as DRAM. The third party generates a SOC design, and provided the third party complies with the set of design rules, the third party SOC design is compatible with the common template mask set. It will be appreciated that, where a mask set does not yet exist to establish an interface for a part from a particular vendor, the common template mask set obviates a requirement for a customer to produce such a mask set. It will be appreciated that this results in substantial savings where the customer desires to use the part from different vendors, and would thus otherwise be required to produce multiple mask sets, generally one per vendor from which the part is sourced. It will be appreciated that while vendor is substantially used herein that the instant application, including the scope of the appended claims, is not meant to be limited thereby. For example, vendor is to be synonymous with merely party, entity or the like. The following description and annexed drawings set forth certain illustrative aspects and implementations. These are indicative of but a few of the various ways in which one or more aspects are employed. Other aspects, advantages, or novel features of the disclosure will become apparent from the following detailed description when considered in conjunction with the annexed drawings. Embodiments or examples, illustrated in the drawings are disclosed below using specific language. It will nevertheless be understood that the embodiments or examples are not intended to be limiting. Any alterations and modifications in the disclosed embodiments, and any further applications of the principles disclosed in this document are contemplated as would normally occur to one of ordinary skill in the pertinent art. It will be appreciated that ‘layer’, as used herein, contemplates a region, and does not necessarily comprise a uniform thickness. For example, a layer is a region, such as an area comprising arbitrary boundaries. A layer is also, for example, a region comprising a variation in thickness. It will be appreciated that for some of the figures herein, one or more boundaries, such as boundary 110A of FIG. 2, for example, are drawn with different heights, widths, perimeters, aspect ratios, etc. relative to one another merely for illustrative purposes, and are not necessarily drawn to scale. For example, because dashed or dotted lines are used to represent different boundaries, if the dashed and dotted lines were drawn on top of one another they would not be distinguishable in the figures, and thus are drawn with different dimensions or slightly apart from one another, in some of the figures, so that they are distinguishable from one another. As another example, where a boundary is associated with an irregular shape, the boundary, such as a box drawn with a dashed line, dotted lined, etc., does not necessarily encompass an entire component in some instances. Conversely, a drawn box does not necessarily encompass merely an associated component, in some instances, but encompasses at least a portion of one or more other components as well. FIG. 1 is a table 100 illustrating an example common template mask set for an electronic article, such as a system on chip (SOC), according to some embodiments. Generally, a third party or any type of consuming entity formulates a SOC design and a first set of masks 140 is used to establish a first interface A (not shown) to connect the SOC to a first part from a first vendor, entity, party, etc. The first set of masks 140 is used to fabricate the first interface A comprising one or more sets of features on the SOC. The first interface A enables the SOC to electrically connect with the first part from the first vendor. As an example, the first set of masks 140 comprises a first mask A 142, a second mask A 144, a third mask A 146, and a fourth mask A 148. The first mask A 142 is used to form a first set of features A on the SOC. Similarly, the second mask A 144, the third mask A 146, and the fourth mask A 148 are used to form a second set of features A, a third set of features A, and a fourth set of features A on the SOC, respectively. Together, the first, second, third, and fourth set of features A form the first interface A on the SOC. A second set of masks 150 are used to fabricate a second interface B (not shown) when a switch is made to a second part from a second vendor, entity, party, etc. It will be appreciated that the first part and the second part correspond to a same part, such as DRAM, for example, but where the first part is provided by the first vendor and the second part is provided by the second vendor. For example, the second set of masks 150 comprises a first mask B 152, a second mask B 154, a third mask B 156, and a fourth mask B 158. The first mask B 152, the second mask B 154, the third mask B 156, and the fourth mask B 158 are used to form a first set of features B, a second set of features B, a third set of features B, and a fourth set of features B on the SOC, respectively. It will be appreciated that the second set of masks 150 is specific to the second vendor and that the first set of masks 140 is specific to the first vendor. In some embodiments, a common template mask set 160 comprises a first mask U 162, a second mask U 164, and a third mask U 166. The first mask U 162 is configured to form the first set of features A and the first set of features B on the SOC. Similarly, the second mask U 164 is configured to form the second set of features A and the second set of features B on the SOC. The third mask U 166 is configured to form the third set of features A and the third set of features B on the SOC. In other words, a ‘U’ mask is functionally equivalent to an ‘A’ mask counterpart and a ‘B’ mask counterpart, at least in part. For example, the first mask U 162 is functionally equivalent to the first mask A 142 and the first mask B 152. In some embodiments, a common template comprises the respective features formed by masks 162, 164, and 166. In other words, the common template comprises the first set of features A, the second set of features A, the third set of features A, the first set of features B, the second set of features B, and the third set of features B. Because masks 162, 164, and 166 of the common template mask set 160 are configured to form features similar to the features associated with masks 142, 152, 144, 154, 146, and 156, the common template mask set 160 is configured to fabricate a common template that is compatible with the first part from the first vendor and the second part from the second vendor. Because the first mask U 162, the second mask U 164, and the third mask U 166 provide features functionally equivalent to the first mask A 142, the first mask B 152, the second mask A 144, the second mask B 154, the third mask A 146, and the third mask B 156, the common template mask set 160 of FIG. 1 reduces a number of masks associated with interface fabrication. A under bump metallization (UBM) mask is used to fabricate one or more vendor specific features on the SOC. For example, the fourth mask A 148 is used to fabricate a first set of UBM features on the SOC when the first vendor is selected. Similarly, the fourth mask B 158 is used to fabricate a second set of UBM features on the SOC when the second vendor is selected. FIG. 2 is a layout view 200 of an example common template mask 130 for an electronic article, such as a system on chip (SOC), according to some embodiments. A common template mask is a combination of a first set of features associated with a first vendor part and a second set of features associated with a second vendor part. For example, 110 is a first set of polygon positions associated with a first vendor part from a first vendor, entity, party, etc. and 120 is a second set of polygon positions associated with a second vendor part from a second vendor, entity, party, etc. Effectively, the sets of polygon positions are design requirements from respective vendors, where polygon positions correspond, for example, to features to be formed in an interface. Generally, a set of polygon positions comprises one or more subsets of polygon positions. A subset of polygon positions comprises one or more patterns. In the example illustrated in FIG. 2, the first set of polygon positions 110 comprises a first subset of polygon positions 114 and a second subset of polygon positions 116. The first subset of polygon positions 114 comprises pattern 110A and pattern 110B. The second subset of polygon positions 116 comprises pattern 102. The second set of polygon positions 120 comprises a third subset of polygon positions 118 and a fourth subset of polygon positions 124. The third subset of polygon positions 118 comprises pattern 120A and pattern 120B. The fourth subset of polygon positions 124 comprises pattern 102. In some embodiments, the first set of polygon positions 110 is functionally equivalent to the second set of polygon positions 120. For example, one or more signals associated with the first set of polygon positions 110, or features ultimately fabricated at the first set of polygon positions 110, are the same as one or more signals associated with the second set of polygon positions 120, or features ultimately fabricated at the second set of polygon positions 120. Because the first set of polygon positions 110 comprises patterns 110A and 110B of the first subset 114 and the second set of polygon positions 120 does not comprise patterns 110A and 110B, the first subset 114 is exclusive to the first set of polygon positions 110. Similarly, because the second set of polygon positions 120 comprises patterns 120A and 120B of the third subset 118 and the first set of polygon positions 110 does not comprise patterns 120A and 120B, the third subset 118 is exclusive to the second set of polygon positions 120. Within the first set of polygon positions 110 and the second set of polygon positions 120, however, the second subset 116 and the fourth subset 124 do overlap. In an example, the overlap occurs because polygons of the second subset 116 and polygons of the fourth subset 124 are positioned according to a commonality, such as an industry standard, for example. While a commonality is not limited to an industry standard, an example industry standard is nevertheless a Joint Electron Devices Engineering Council (JEDEC) standard. The common template mask 130 of FIG. 1 is generated based on the first set of polygon positions 110 and the second set of polygon positions 120, as illustrated by arrows pointing from 110 and 120 to 130. For example, the common template mask 130 comprises the first subset 114 of the first set of polygon positions 110, the third subset 118 of the second set of polygon positions 120, and the second subset 116 of the first set of polygon positions 110 or the fourth subset 124 of the second set of polygon positions. Explained in another way, the common template mask 130 is generated based on a superset or a union of the first set of polygon positions 110 and the second set of polygon positions 120. Because the second subset 116 and the fourth subset 124 share overlapping positions, the second subset 116 and the fourth subset 124 are merged in the common template mask 130, illustrated as 116. Since the common template mask 130 combines the first set of polygon positions 110 with the second set of polygon positions 120, the common template mask 130 facilitates formation of features that satisfy design requirements for both the first vendor and the second vendor. Because of this, the common template formed by the common template mask 130 is compatible with parts from both the first vendor and the second vendor. In this way, the common template mask 130 comprises a vendor neutral design with regard to the first vendor and the second vendor. It will be appreciated that separate masks associated with the first set of polygon positions 110, such as the first mask A 142 of FIG. 1, and the second set of polygon positions 120, such as the first mask B 152 of FIG. 1, are not required when using the common template mask 130. Accordingly, a number of masks associated with fabricating a SOC design, or an interface associated therewith, is mitigated. Accordingly, a number of manufacturing process variables associated with retooling for multiple masks is reduced as well. FIG. 3 is a flow diagram of an example method 300 for incorporating a common template into an electronic article design, such as a system on chip (SOC) design, according to some embodiments. A common template mask set is generated from a superset of design requirements received from multiple vendors, entities, parties, party, etc. Design requirements are generally expressed as polygon positions for a mask, or rather features of an interface fabricated using the mask. At 302, a first set of polygon positions is received. The first set of polygon positions is a set of design requirements associated with a first part from a first vendor. At 304, a second set of polygon positions is received from a second vendor. The second set of polygon positions is a set of design requirements associated with a second part from a second vendor. At 306, a common template mask set comprising one or more common template masks is generated based on the first set of polygon positions and the second set of polygon positions. In some embodiments, a common template mask of the common template mask set is generated based on a superset or union of the first set of polygon positions and the second set of polygon positions. At a later stage, a common template mask of the common template mask set is used to fabricate a portion of a common template, such as a set of features on a SOC. In some embodiments, one or more UBM masks are generated for the respective vendors. For example, in some instances, there are no common features among vendors and thus a vendor specific UBM mask is required for different vendors. Accordingly, at 308, a first under bump metallization (UBM) mask or a second UBM mask is generated based on the first set of polygon positions or the second set of polygon positions, respectively. At a later stage, a UBM mask is used to fabricate micro-bumps on a SOC. The micro-bumps enable the SOC to be electrically connected to a first part from a first vendor or a second part from a second vendor, for example. At 310, a set of design rules is generated based on the common template mask set and provided to a third party, such as a customer. The set of design rules enables an SOC designer to create a design that can be fabricated utilizing the common template mask set. The set of design rules are indicative of suggested protocol to be followed during SOC design in order for the common template mask set to be used in conjunction with the SOC. For example, when a SOC or a SOC design in accordance with the set of design rules is received, the common template masks of the common template mask set are used to fabricate a common template on the SOC. Because the common template mask set is being used, the common template is compatible with the first part from the first vendor and the second part from the second vendor. In some embodiments, a first design rule defines a position associated with an input-output (I/O) connection for the SOC. For example, an I/O connection is a micro-bump connection or a through silicon via (TSV) connection. A second design rule defines a number of I/O connections at a top layer of the SOC. Explained in another way, a design rule facilitates mating between a third party SOC and a common template mask of a common template mask set to fabricate a common template on the third party SOC. In other words, because the design rule is indicative of a suggested connection between third party logic of the third party SOC and an interface, such as a common template, a SOC designed according to the design rule enables one to fabricate the interface on the SOC using the common template mask set, rather than a custom mask set. It will be appreciated that some SOCs are fabricated active side up, while other SOCs are fabricated active side down, as will be described in FIG. 5 and FIG. 6, respectively. A third design rule associated with an active side up SOC design defines a position associated with a micro-bump for the SOC. A fourth design rule associated with an active side down SOC design defines a position associated with a through silicon via (TSV) for the SOC. FIG. 4 is a flow diagram of an example method 400 for incorporating a common template into an electronic article design, such as a system on chip (SOC) design, according to some embodiments. When a set of design rules generated based on a common template mask set is provided to a third party, the third party generates a third party SOC design in accordance with the set of design rules. In some embodiments, the third party SOC design that is in accordance with the set of design rules is received at 402. In some embodiments, a third party SOC is fabricated based on the third party SOC design at 402. In other embodiments, a fabricated third party SOC is received at 402. That is, rather than fabricating the SOC based upon the third part SOC design, the third party SOC is already fabricated, according to the third party SOC design, such as by a different entity, for example, and is merely received at 402. Because the set of design rules is based on the common template mask set, a third party following the set of design rules is not required to develop a custom set of masks to fabricate an interface on the SOC. Thus, having a third party use design rules that are in conformance with a common template mask set allows the third party to generate a third party SOC design that can be satisfied by multiple vendors, without requiring multiple mask sets specific to each vendor. During an intermediate fabrication stage at 404, a common template is fabricated on the third party SOC using the common template mask set. A third party SOC design often comprises a re-distribution layer (RDL) on a backside or a non-active region of the SOC. In some embodiments, the common template is fabricated on a backside of the third party SOC using the common template mask set. The common template mask set comprises one or more common template masks. For example, a common template mask of the common template mask set is configured to form a first set of features associated with a first part for a first vendor, such as a first set of features A associated with the first mask A 142 of FIG. 1, and a second set of features associated with a second part for a second vendor, such as the first set of features B associated with the first mask B 152 of FIG. 1. Together, the common template masks of the common template mask set form one or more sets of features associated with the first part from the first vendor and one or more sets of features associated with the second part from the second vendor. Effectively, the common template masks of the common template mask set form a common template that comprises the respective features. As discussed with regard to FIG. 2, a common template formed by a common template mask set comprising the common template mask 130 is compatible with a part from the first vendor as well when the part is sourced from the second vendor. In this way, the common template mask set is used to create a ‘standard’ interface on the SOC. In other words, the common template has a vendor neutral design. At 406, a under bump metallization (UBM) layer is fabricated on the SOC based on a vendor selection of a part, such as DRAM, for the SOC. For example, the vendor selection comprises a first vendor selection or a second vendor selection. Up to this stage, the SOC, the common template fabricated on the SOC, and associated fabrication processing, such as common template masks used, have been vendor independent. For example, with reference to FIG. 1, the common template mask set 160 comprising the first mask U 162, the second mask U 164, and the third mask U 166 are used to fabricate a common template comprising sets of features associated with the first mask A 142, the second mask A 144, the third mask A 146, the first mask B 152, the second mask B 154, and the third mask B 156. The UBM layer, however, is vendor specific. In other words, the fourth mask A 148 is used to fabricate a first UBM layer when a part from the first vendor is selected, while the fourth mask B 158 is used to fabricate a second UBM layer when a part from a second vendor is selected. In some embodiments, the UBM layer comprises micro-bumps that are used to electrically connect the SOC to a part from a vendor. For example, if the first vendor is selected, the fourth mask A 148 is used to fabricate a first UBM layer comprising one or more micro-bumps that are configured to connect the SOC to the part from the first vendor. Similarly, if the second vendor is selected, the fourth mask B 158 is used to fabricate a second UBM layer comprising one or more micro-bumps that are configured to connect the SOC to the part from the second vendor. Explained in another way, when a part from a vendor is integrated into a SOC design, a UBM mask corresponding to the vendor is used to fabricate a UBM layer to connect the part from the vendor to the SOC. In this way, merely a UBM mask, such as the fourth mask A 148 or the fourth mask B 158, is changed when a different vendor is selected to provide a part for a SOC. Because the common template mask set is compatible with parts from multiple vendors, vendor selection is not required at an earlier stage, such as fabrication of the common template. Since the common template mask set is adaptable to multiple vendors, no mask redesign is necessary for the common template when the third party changes a vendor selection. Further, if a third party switches vendors for a part, merely the UBM mask is changed, resulting in a reduced amount of re-tooling or setup when the third party changes vendors. FIG. 5 is a cross-sectional view 500 of a portion of an active side up electronic article design, such as an active side up system on chip (SOC) design, according to some embodiments. A common template 582 is fabricated on the SOC 570 based on a common template mask set. In some embodiments, such as with an active side up SOC design, the common template 582 comprises a first dielectric region 410, a second dielectric region 420, a first metal region 412 within the first dielectric region 410, a second metal region 422 within the second dielectric region 420, and an interconnect 402. A design rule associated with a common template mask of the common template mask set defines a position associated with a micro-bump at 430. When a third party SOC design follows this design rule by positioning a micro-bump at this position, a common template mask set can be used to fabricate the common template 582 on the SOC 570. The common template 582 facilitates connecting the SOC 570 to a vendor part, such as DRAM 590 after a vendor specific UBM layer 580 is fabricated on the common template 582. FIG. 6 is a cross-sectional view 600 of a portion of an active side down electronic article design, such as an active side down system on chip (SOC) design, according to some embodiments. A common template 584 is fabricated on the SOC 570 based on a common template mask set. The common template 584 comprises a first dielectric region 410, a second dielectric region 420, a first metal region 412 within the first dielectric region 410, a second metal region 422 within the second dielectric region 420, an interconnect 402, a silicon region 540, a through silicon via (TSV) 502, a passivation region 550, and a backside metal region 552. A design rule associated with a common template mask defines a position associated with the TSV at 530. When a third party SOC design follows this design rule by placing a TSV line at a predetermined or fixed position, such as at 530, a common template mask set can be used to fabricate the common template 584 on the SOC 570. The common template 584 connects the SOC 570 to a vendor part, such as DRAM 590 after a vendor specific UBM layer 560 is fabricated on the common template 584. Still another embodiment involves a computer-readable medium comprising processor-executable instructions configured to implement one or more of the techniques presented herein. An example embodiment of a computer-readable medium or a computer-readable device is illustrated in FIG. 7, wherein an implementation 700 comprises a computer-readable medium 708, such as a CD-R, DVD-R, flash drive, a platter of a hard disk drive, etc., on which is encoded computer-readable data 706. This computer-readable data 706, such as binary data comprising a plurality of zero's and one's as shown in 706, in turn comprises a set of computer instructions 704 configured to operate according to one or more of the principles set forth herein. In one such embodiment 700, the processor-executable computer instructions 704 are configured to perform a method 702, such as at least some of the exemplary method 300 of FIG. 3 or at least some of exemplary method 400 of FIG. 4. In another embodiment, the processor-executable instructions 704 are configured to implement a system. Many such computer-readable media are devised by those of ordinary skill in the art that are configured to operate in accordance with the techniques presented herein. As used in this application, the terms “component”, “module,” “system”, “interface”, and the like are generally intended to refer to a computer-related entity, either hardware, a combination of hardware and software, software, or software in execution. For example, a component may be, but is not limited to being, a process running on a processor, a processor, an object, an executable, a thread of execution, a program, or a computer. By way of illustration, both an application running on a controller and the controller can be a component. One or more components residing within a process or thread of execution and a component may be localized on one computer or distributed between two or more computers. Furthermore, the claimed subject matter is implemented as a method, apparatus, or article of manufacture using standard programming or engineering techniques to produce software, firmware, hardware, or any combination thereof to control a computer to implement the disclosed subject matter. The term “article of manufacture” as used herein is intended to encompass a computer program accessible from any computer-readable device, carrier, or media. Of course, many modifications may be made to this configuration without departing from the scope or spirit of the claimed subject matter. FIG. 8 and the following discussion provide a description of a suitable computing environment to implement embodiments of one or more of the provisions set forth herein. The operating environment of FIG. 8 is only one example of a suitable operating environment and is not intended to suggest any limitation as to the scope of use or functionality of the operating environment. Example computing devices include, but are not limited to, personal computers, server computers, hand-held or laptop devices, mobile devices, such as mobile phones, Personal Digital Assistants (PDAs), media players, and the like, multiprocessor systems, consumer electronics, mini computers, mainframe computers, distributed computing environments that include any of the above systems or devices, and the like. Generally, embodiments are described in the general context of “computer readable instructions” being executed by one or more computing devices. Computer readable instructions are distributed via computer readable media as will be discussed below. Computer readable instructions are implemented as program modules, such as functions, objects, Application Programming Interfaces (APIs), data structures, and the like, that perform particular tasks or implement particular abstract data types. Typically, the functionality of the computer readable instructions are combined or distributed as desired in various environments. FIG. 8 illustrates an example of a system 800 comprising a computing device 812 configured to implement one or more embodiments provided herein. In one configuration, computing device 812 includes at least one processing unit 816 and memory 818. Depending on the exact configuration and type of computing device, memory 818 may be volatile, such as RAM, non-volatile, such as ROM, flash memory, etc., or some combination of the two. This configuration is illustrated in FIG. 8 by dashed line 814. In other embodiments, device 812 includes additional features or functionality. For example, device 812 also includes additional storage such as removable storage or non-removable storage, including, but not limited to, magnetic storage, optical storage, and the like. Such additional storage is illustrated in FIG. 8 by storage 820. In some embodiments, computer readable instructions to implement one or more embodiments provided herein are in storage 820. Storage 820 also stores other computer readable instructions to implement an operating system, an application program, and the like. Computer readable instructions are loaded in memory 818 for execution by processing unit 816. The term “computer readable media” as used herein includes computer storage media. Computer storage media includes volatile and nonvolatile, removable and non-removable media implemented in any method or technology for storage of information such as computer readable instructions or other data. Memory 818 and storage 820 are examples of computer storage media. Computer storage media includes, but is not limited to, RAM, ROM, EEPROM, flash memory or other memory technology, CD-ROM, Digital Versatile Disks (DVDs) or other optical storage, magnetic cassettes, magnetic tape, magnetic disk storage or other magnetic storage devices, or any other medium which can be used to store the desired information and which can be accessed by device 812. Any such computer storage media is part of device 812. The term “computer readable media” includes communication media. Communication media typically embodies computer readable instructions or other data in a “modulated data signal” such as a carrier wave or other transport mechanism and includes any information delivery media. The term “modulated data signal” includes a signal that has one or more of its characteristics set or changed in such a manner as to encode information in the signal. Device 812 includes input device(s) 824 such as keyboard, mouse, pen, voice input device, touch input device, infrared cameras, video input devices, or any other input device. Output device(s) 822 such as one or more displays, speakers, printers, or any other output device are also included in device 812. Input device(s) 824 and output device(s) 822 are connected to device 812 via a wired connection, wireless connection, or any combination thereof. In some embodiments, an input device or an output device from another computing device are used as input device(s) 824 or output device(s) 822 for computing device 812. Device 812 also includes communication connection(s) 826 to facilitate communications with one or more other devices. According to some aspects, a method for incorporating a common template into a system on chip (SOC) design is provided, comprising receiving a first set of polygon positions from a first vendor. Additionally, the method comprises receiving a second set of polygon positions from a second vendor. The method comprises generating a common template mask set based on the first set of polygon positions and the second set of polygon positions. The method comprises generating a first under bump metallization (UBM) mask for the first vendor or a second UBM mask for the second vendor based on the first set of polygon positions or the second set of polygon positions. The method comprises generating a set of design rules based on the common template mask set, a first design rule defining a position associated with an input-output (I/O) connection for a system on chip (SOC). According to some aspects, a method for incorporating a common template into a system on chip (SOC) design is provided, comprising receiving a third party system on chip (SOC) design that is in accordance with a set of design rules based on a common template mask set. Additionally, the method comprises fabricating a common template in conjunction with the third party SOC design based on the common template mask set. According to some aspects, common template mask for a system on chip (SOC) is provided, comprising a first pattern, a second pattern, and a third pattern. For example, the first pattern is associated with a first subset of a first set of polygon positions from a first vendor. For example, the second pattern is associated with a first subset of a second set of polygon positions from a second vendor. For example, the third pattern is associated with a second subset of the first set of polygon positions from the first vendor that overlaps with a second subset of the second set of polygon positions from the second vendor. Although the subject matter has been described in language specific to structural features or methodological acts, it is to be understood that the subject matter of the appended claims is not necessarily limited to the specific features or acts described above. Rather, the specific features and acts described above are disclosed as example forms of implementing the claims. Various operations of embodiments are provided herein. The order in which some or all of the operations are described should not be construed as to imply that these operations are necessarily order dependent. Alternative ordering will be appreciated based on this description. Further, it will be understood that not all operations are necessarily present in each embodiment provided herein. Further, unless specified otherwise, “first,” “second,” or the like are not intended to imply a temporal aspect, a spatial aspect, an ordering, etc. Rather, such terms are merely used as identifiers, names, etc. for features, elements, items, etc. For example, a first channel and a second channel generally correspond to channel A and channel B or two different or identical channels. Moreover, “exemplary” is used herein to mean serving as an example, instance, illustration, etc., and not necessarily as advantageous. As used in this application, “or” is intended to mean an inclusive “or” rather than an exclusive “or”. In addition, “a” and “an” as used in this application are generally construed to mean “one or more” unless specified otherwise or clear from context to be directed to a singular form. Also, at least one of A and B and/or the like generally means A or B or both A and B. Furthermore, to the extent that “includes”, “having”, “has”, “with”, or variants thereof are used in either the detailed description or the claims, such terms are intended to be inclusive in a manner similar to the term “comprising”. Also, although the disclosure has been shown and described with respect to one or more implementations, equivalent alterations and modifications will occur based on a reading and understanding of this specification and the annexed drawings. The disclosure includes all such modifications and alterations and is limited only by the scope of the following claims. |
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047175320 | summary | BACKGROUND OF THE INVENTION The present invention relates to a pressurized water nuclear reactor power plant and specifically to an improved sparger system for use in the pressurizer relief tank for use in such plants which protects the plant from overpressure and thus, provides safe plant operation. In a pressurized water nuclear reactor plant, a primary coolant loop and secondary coolant loop are used to produce steam for the production of electricity. In the primary coolant loop, a pressurized fluid is passed through a nuclear reactor and, after being heated, through a line which contains a pressurizer, to a steam generator. The heated fluid enters the primary side of the steam generator which is divided into an inlet section and an outlet section by a divider plate. A tube sheet divides the steam generator into the primary side and a second side, which tube sheet has an array of holes having U-shaped heat transfer tubes inserted therein, which communicate between the inlet section and outlet section of the primary side of the steam generator. In operation, the heat pressurized fluid passes through the U-shaped heat transfer tubes and is discharged from the outlet section of the primary side of the steam generator to a line, containing a primary coolant pump, back to the reactor in a continuous closed loop. Secondary coolant is passed through the secondary side of the steam generator where it is converted into steam by heat released by the primary coolant passing through the U-shaped heat transfer tubes, which steam is used to drive a turbine to produce electricity. The reactor coolant pressure is controlled by a pressure control system containing a pressurizer, which is a vertical, cylindrical vessel with hemispherical top and bottom heads, wherein water and steam are maintained in equilibrium by electrical heaters and water sprays. Steam can be formed by activating the heaters to increase the pressure in the primary coolant loop, or condensed by the water sprays to reduce the pressure. Power operated relieve valves and spring loaded safety valves are connected to the pressurizer and discharge to a pressurizer relief tank, where steam from the pressurizer is condensed and cooled by mixing with water. The pressure control system for the primary loop thus includes the pressurizer and the associated sprays, heaters, power operated relief valves, safety valves, relief tank, and surge lines. This equipment is designed to accommodate changes in system volume and to limit changes in system pressure due to reactor coolant loop temperature variations during all modes of plant operation. To reduce the problem of leakage through the valve seats, a water seal is maintained upstream of each valve seat of the pressurizer valves. The pipes connecting the pressurizer nozzles to their respective valves are shaped in the form of a loop seal. If the pressurizer pressure exceeds the set pressure of the valves, they will open, and the water from the loop seal will discharge during the accumulation period. In order to avoid steam release to the containment from the nuclear reactor during normal and upset condition operation, the power operated relief valves (PORV) and the safety valves (SV) are thus mounted on the pressurizer and are routed to a pressurizer relief tank. The pressurizer relief tank is a partially water filled tank having a nitrogen cover gas. Hot fluids, water and steam, discharged from the power operated relief valves and safety valves are distributed into the water in the pressurizer relief tank by means of a sparger therein which conventionally is comprised of a straight conduit having a plurality of orifices therein. Typically, the sparger comprises a 12" pipe with a plurality of small orifices (about 1/2 inch diameter) in long rows no more than about 30.degree. above and below the horizontal axis of the pipe. The number of orifices in the sparger is based on the efficient condensation of the steam flow through the sparger associated with the peak volumetric surge rate into the pressurizer during an assumed loss of load (100%) without immediate reactor trip or shutdown. Long, approximately 1" in diameter, steam plumes will be produced through each orifice in the sparger for this basic design event. For lower steam flowrates, steam could bubble out of each orifice, while for higher steam flowrates, the steam plumes, almost jets, could penetrate the water surface. For off-design flowrates, condensation efficiency of the steam in the water is reduced. The pressure drop across the sparger is part of the total safety valve discharge system drop which is conventionally limited to a 500 pounds per square inch differential due to the limitations on the valve backpressure compensating bellows. The 500 pounds per square inch gauge bellows limit, conventionally used, is more conservative than the maximum backpressure of 750 pounds per square inch gauge normally permitted. The design of a power operated relief valve-safety valve system that accommodates for an anticipated transient without trip (ATWT) is desired. An anticipated transient without trip is an event in which the diverse and independent reactor trip, or shutdown, safety systems fail to immediately shutdown the reactor for an extended time period. Such an event would result in a delay of the reactor trip until after the pressurizer had become completely filled with coolant water, or water solid. Following such an event, the reactor will eventually be shutdown either automatically or manually, but a delay is present. In a typical ATWT event, the insurge of reactor coolant into the pressurizer continues until and after the pressurizer is filled with water. The power operated relief valves open and relieve steam and safety valves open with the pressurizer filled with water. The safety valves discharge water for about 50 seconds while the power operated relief valves discharge water for about 120 seconds. As a result, the pressure drop across the discharge piping and sparger increases dramatically. Valve backpressure increases to approximately 1000 pounds per square inch gauge, well above both the bellows and standard test limits. In orde to accommodate an anticipated transient without trip event, modification of existing equipment was required. According to the present invention, the discharge piping was increased to 16 inch diameter piping from standard 12" piping, and sparger modifications are provided. A simple increase in the size of the orifices in the sparger was not acceptable since steam condensation efficiency for the nominal (loss of load) design basis was insufficient. Inefficient condensation causes premature rupture of the overpressure rupture disks on the pressurizer relief tank and unnecessary discharge of such steam to the containment. It is an object of the present invention to provide a pressure control system for a pressurized water nuclear reactor plant that will accommodate normal and anticipated transient without trip events without unnecessary discharge of steam to the containment. It is another object of the present invention to provide an improved sparger for use in a pressurizer relief tank that provides efficient steam distribution and condensation in the pressurizer relief tank while limiting the pressure drop for normal and anticipated transient without trip events. SUMMARY OF THE INVENTION A pressure control system for a pressurized water nuclear reactor plant that includes a pressurizer and safety and relief valves which discharge to a pressurizer relief tank having a two stage sparger in the pressurizer relief tank that is comprised of a primary conduit having orifices, a secondary conduit having orifices, and an interconnecting valve means that is responsive to a predetermined pressure differential between the two conduits. The check valve is preferably a spring biased valve and the primary conduit preferably larger in diameter than the secondary conduit. A preferred construction of the secondary conduit is one which has a bifurcated construction with two leg sections extending back towards the primary conduit, which provides for a compact design. The orifices in the secondary conduit, or at least a portion thereof, are located in the secondary conduit such that they are at a lower level in the pressurizer relief tank than the orifices of the primary conduit. |
description | This application claims the benefit under 35 U.S.C. § 119(e) of U.S. Provisional Patent Application Ser. No. 62/610,395, filed Dec. 26, 2017, the disclosure of which is hereby incorporated herein in its entirety by this reference. This invention was made with government support under Contract Number DE-AC07-05-1D14517 awarded by the United States Department of Energy. The government has certain rights in the invention. Embodiments of the disclosure relate generally to methods of additively manufacturing one or more structures of a nuclear reactor. More particularly, embodiments of the disclosure relate to methods and apparatuses for manufacturing a porous nuclear fuel by additive manufacturing, to nuclear fuel reactors or components thereof including the porous nuclear fuel, and to related methods. Nuclear reactors include fuel rods, plates, or assemblies containing a nuclear fuel surrounded by cladding. Fission of the nuclear fuel produces heat, which in turn, is used to generate electricity, such as by powering a turbine with steam generated by the heat. Nuclear fuels include ceramic fuels, or metallic and cermet fuels. Metallic and cermet fuels are often preferred to conventional ceramic fuels because the metallic and cermet fuels exhibit a greater thermal conductivity than conventional fuels including only ceramic fuel materials. In use and operation, the nuclear fuel may exhibit so-called “neutron-induced swelling” wherein the nuclear fuel increases in volume and decreases in density when subjected to intense neutron radiation. In addition to neutron-induced swelling, as a result of thermal expansion, the nuclear fuel increases in volume responsive to exposure to the elevated temperatures of the nuclear reactor. As the nuclear fuel swells, the nuclear fuel may impede on structures disposed around the nuclear fuel, such as cladding containing the nuclear fuel. Swelling of the nuclear fuel results in undesired stresses on the cladding surrounding and contacted by the swollen nuclear fuel. If the expansion of nuclear fuel is more than a certain amount, the cladding surrounding the nuclear fuel may crack or otherwise fail. In order to accommodate swelling of the nuclear fuel under operating conditions, it has been known to dispose molten sodium between the cladding and the nuclear fuel. The molten sodium is a displaceable media that bonds the cladding material to the nuclear fuel. However, as the nuclear fuel expands toward the cladding, the sodium is pushed into a head spaced of the fuel rod assembly, reducing the heat transfer from the nuclear fuel to the cladding, ultimately reducing the effectiveness of the fuel rod. In addition, the sodium is exposed to fission products and must be treated prior to disposal thereof after the useful like of the fuel rod. Further, in the case of light water reactors, sodium is typically undesired due to the reaction between water and sodium. Nuclear fuels are conventionally coupled to a heat exchange mechanism for transferring thermal energy from the nuclear fuel to another portion of the nuclear reactor for power generation. Without the use of subtractive machining, the nuclear fuel is conventionally limited to right cylindrical geometries. However, the geometry constraints of the nuclear fuel formed by conventional methods may limit the manner in which the nuclear fuel mechanically interfaces with heat exchangers or power conversion mechanisms. Embodiments disclosed herein include methods of additively manufacturing structures for a nuclear reactor, and to related nuclear fuels, fuel rods, and structures of a nuclear reactor. For example, in accordance with one embodiment, a method of forming a fuel rod for a nuclear reactor comprises disposing a powder comprising particles of a fuel material on a substrate, exposing the powder to energy from an energy source to form a first layer of a nuclear fuel, the first layer comprising inter-granular bonds between the particles of the fuel material, disposing additional powder comprising particles of the fuel material over the first layer of the nuclear fuel, and exposing the additional powder to energy from the energy source to form a second layer of the nuclear fuel and to form the nuclear fuel to have a void fraction greater than about 0.20, the second layer comprising inter-granular bonds between the additional powder and the first layer of the nuclear fuel. In additional embodiments, a fuel rod comprising a porous nuclear fuel comprises a nuclear fuel having a porous structure having void fraction greater than about 0.10, and cladding disposed around the nuclear fuel. Illustrations presented herein are not meant to be actual views of any particular material, component, or system, but are merely idealized representations that are employed to describe embodiments of the disclosure. The following description provides specific details, such as material types, dimensions, and processing conditions in order to provide a thorough description of embodiments of the disclosure. However, a person of ordinary skill in the art will understand that the embodiments of the disclosure may be practiced without employing these specific details. Indeed, the embodiments of the disclosure may be practiced in conjunction with conventional fabrication techniques employed in the industry. In addition, the description provided below does not form a complete process flow, apparatus, or system for forming a nuclear fuel element, a component of a nuclear reactor core, another structure, or related methods. Only those process acts and structures necessary to understand the embodiments of the disclosure are described in detail below. Additional acts to form a nuclear fuel element, a component of a nuclear reactor core, or another structure may be performed by conventional techniques. Further, any drawings accompanying the present application are for illustrative purposes only and, thus, are not drawn to scale. Additionally, elements common between figures may retain the same numerical designation. As used herein, the term “metallic foam” means and includes a material comprising a metal component and exhibiting a void fraction (e.g., porosity) greater than about 0.10. Accordingly, metallic foam materials may not exhibit a full theoretical density. According to embodiments described herein, a structure comprising one or more components of a nuclear reactor (e.g., one or more components of a nuclear reactor core, a nuclear fuel, cladding, a fuel rod, heat exchanger mechanisms associated with a nuclear reactor core, etc.) is formed layer by layer in an additive manufacturing process, which may also be characterized as a direct material deposition process. Since the structure is formed layer by layer, the structure may be formed to exhibit complex cross-sectional geometries, non-uniform external surface topographies, and compositional features that are unobtainable or difficult to manufacture according to conventional methods. In some embodiments, the structure is formed to exhibit a desired porosity, and a desired shape and size. In some embodiments, a nuclear fuel is formed to exhibit a porosity, which may facilitate accommodation of expansion (e.g., swelling) of the nuclear fuel during use and operation thereof in a nuclear reactor. In addition, the nuclear fuel may be formed to have a desired geometry, such as, for example, internal cooling channels, channels for coupling the nuclear fuel to a heat exchanger mechanism, etc. Since the nuclear fuel is formed layer by layer, the nuclear fuel may be formed to exhibit a compositional gradient along a length thereof (e.g., one or more layers of the nuclear fuel may exhibit a different composition than one or more other layers of the nuclear fuel), a compositional gradient along a radius thereof, or a combination thereof. In some embodiments, the nuclear fuel may include one or more dopants therein to increase a tensile strength of the nuclear fuel. According to further embodiments, a fuel rod comprising the nuclear fuel and a cladding material may be formed by additive manufacturing. In some embodiments, the fuel rod may include a fission barrier, a reactor poison, or both. Since the fuel rod is formed by additive manufacturing, the fuel rod may be formed to exhibit a desired internal and external geometry. In some embodiments, a cross-sectional shape of the fuel rod may be selected to increase a thermal conductivity of the nuclear fuel to the cladding material. In some embodiments, the fuel rod may be formed to include channels in the nuclear fuel structure, the channels configured to receive a heat transfer fluid therein (e.g., air, water, etc.). In some embodiments, the fuel rod may be formed to exhibit a compositional gradient along a length thereof, along a radius thereof, or a combination thereof. In further embodiments, a portion of a reactor core may be formed by additive manufacturing. After fabrication of the structure, at least a portion of the structure may be exposed to annealing conditions (e.g., hot isostatic pressing, spark plasma sintering, or one or more other densification processes) to densify at least a portion of the structure. In some embodiments, exposing the structure to annealing conditions may densify portions of the structure (e.g., cladding), while other portions thereof (e.g., the nuclear fuel) are substantially unaffected and not densified (e.g., remain porous). Since the cladding may exhibit a lower melting temperature than the nuclear fuel, the nuclear fuel may not be affected by exposure of the structure to the annealing conditions. In some such embodiments, the nuclear fuel may remain at a theoretical density less than a predetermined amount (e.g., less than about 90% theoretical density, less than about 80% theoretical density, less than about 70% theoretical density) while other portions of the structure (e.g., the cladding material) are densified to full density, a theoretical density greater than about 80%, greater than about 90%, greater than about 95%, greater than about 98%, or even greater than about 99%. Referring to FIG. 1, a system 100 for additively manufacturing one or more components of a nuclear reactor is illustrated, in accordance with embodiments of the disclosure. The system 100 may be used to additively manufacture, for example, a nuclear fuel, a nuclear fuel surrounded by cladding, a fuel rod, other components of a nuclear reactor, or combinations thereof. The system 100 includes a powder feed 102 comprising sources of one or more powder constituents used to form a product to be additively manufactured. The powder feed 102 may comprise particles of a nuclear fuel material, particles of a cladding material, particles of a nuclear reactor poison, particles of a fission barrier material, one or more dopants, particles of a heat exchange mechanism (e.g., particles of a heat pipe), another component of a nuclear reactor (e.g., particles making up one or more components of a nuclear reactor core), or combinations thereof. By way of nonlimiting example, where the powder feed 102 comprises a nuclear fuel, the powder feed may include particles of uranium, zirconium, tungsten, tantalum, iridium, uranium dioxide (UO2), uranium oxide (e.g., U3O8), uranium nitride (e.g., UN, U2N3, etc.), uranium borides (e.g., UB2, UB4), a transuranic material (e.g., plutonium, plutonium oxide), thorium, oxides thereof, another nuclear fuel material, or combinations thereof. In some embodiments, the powder feed 102 comprises a mixture of uranium and at least one of zirconium, molybdenum, and tungsten. By way of nonlimiting example, the powder feed 102 may comprise uranium and zirconium and may include between about 1.0 weight percent and about 15.0 weight percent uranium (e.g., between about 1.0 weight percent and about 5.0 weight percent, between about 5.0 weight percent and about 10.0 weight percent, or between about 10.0 weight percent and about 15.0 weight percent uranium) and between about 85.0 weight percent and about 99.0 weight percent zirconium (e.g., between about 85.0 weight percent and about 90.0 weight percent, between about 90.0 weight percent and about 95.0 weight percent, or between about 95.0 weight percent and about 99.0 weight percent zirconium). In some embodiments, the powder feed 102 may comprise about 10.0 weight percent uranium and about 90.0 weight percent zirconium to form a nuclear fuel comprising about 10.0 weight percent uranium and about 90.0 weight percent zirconium which may be referred to as U-10Zr fuel. Where the powder feed 102 includes particles of a cladding material, the powder feed 102 may include particles of zirconium, a stainless steel alloy (e.g., 316 stainless steel), nickel, iron, chromium, molybdenum, titanium, tungsten, or combinations thereof. Where the powder feed 102 includes particles of a fission barrier material, the powder feed 102 may include particles of zirconium, vanadium, another material, or combinations thereof. In some embodiments, the powder feed 102 includes one or more dopants with which a nuclear fuel material may be mixed. By way of nonlimiting example, the dopants may include a metal oxide (aluminum oxide, zirconium oxide, etc.), carbon nanotubes, carbon nanotubes coated with a metal oxide (e.g., aluminum oxide, zirconium oxide, another metal oxide, or combinations thereof), another material, or combinations thereof. In some embodiments, the dopant may facilitate improved tensile strength of a structure formed from the powder feed 102. In some embodiments, coating carbon nanotubes with a metal oxide may improve a wetting angle of the coated carbon nanotubes and may improve emulsification of the coated carbon nanotubes in a metal phase during sintering of the carbon nanotubes in a metal fuel network. As will be described herein, the one or more dopants may increase a tensile strength of a nuclear fuel including the one or more dopants. Of course, the powder feed 102 may include other materials. In some embodiments, the powder feed 102 includes one or more burnable poison materials, such as boron, gadolinium, Gd2O3, B4C, etc., another material exhibiting a high thermal neutron absorption cross-section, and combinations thereof. In other embodiments, the powder feed 102 includes poisons such as krypton, molybdenum, neodymium, hafnium, another neutron absorber, or combinations thereof. In some embodiments, the powder feed 102 includes at least some particles of a nuclear fuel (e.g., uranium, uranium oxide, uranium dioxide, uranium nitrides, uranium borides, a transuranic material, etc.) coated with a layer of the burnable poison. In yet other embodiments, the powder feed 102 includes one or more materials for forming a neutron reflector. In some such embodiments, the powder feed 102 includes, for example, particles of beryllium, particles of graphite, another material exhibiting a sufficient neutron reflectivity, or combinations thereof. In some embodiments, the powder feed 102 is in fluid communication with a powder delivery nozzle 104. The powder feed 102 may be provided to the powder delivery nozzle 104 as a mixture having a desired composition. In other embodiments, the powder may be provided to the powder delivery nozzle 104 as separate components (e.g., zirconium and uranium) that are mixed at the powder delivery nozzle 104. The powder delivery nozzle 104 may be positioned and configured to deliver the powder feed 102 to a surface of a substrate 106 on which a structure 108 is formed. The powder delivery nozzle 104 may be configured to deliver more than one powder feed 102 composition to the substrate 106 concurrently. In other words, the powder delivery nozzle 104 may be in fluid communication with powders having more than one composition and may be used to form the structure 108 having one or more different composition therethrough. Accordingly, although only one powder delivery nozzle 104 is illustrated in FIG. 1, in some embodiments, the system 100 includes more than one powder delivery nozzle 104, each powder delivery nozzle 104 in fluid communication with a powder feed 102 having a different composition than the other powder delivery nozzles 104. By way of nonlimiting example, in some embodiments, the system 100 includes a powder delivery nozzle 104 in fluid communication with a powder feed 102 comprising a nuclear fuel material, a powder delivery nozzle 104 in fluid communication with a powder feed 102 comprising a cladding material, a powder delivery nozzle 104 in fluid communication with a powder feed 102 comprising a fission barrier material, and a powder delivery nozzle 104 in fluid communication with a powder feed 102 comprising a poison material. In other embodiments, the powder delivery nozzle 104 may be in fluid communication with a plurality of powder feed 102 materials. In some such embodiments, the powder delivery nozzle 104 is configured to receive powder from different powder feed 102 materials and configured to dispose powders of different compositions on the substrate 106. The substrate 106 and the structure 108 are disposed on a table 110, which may comprise, for example, a triaxial numerical control machine. Accordingly, the table 110 may be configured to move along at least three axes. By way of nonlimiting example, the table 110 may be configured to move in the x-direction (i.e., left and right in the view illustrated in FIG. 1), the y-direction (i.e., into and out of the page in the view illustrated in FIG. 1), and the z-direction (i.e., up and down in the view illustrated in FIG. 1). The table 110 may be operably coupled with a central processing unit 112 configured to control the table 110. In other words, movement of the table 110 may be controlled through the central processing unit 112, which may comprise a control program for a processor including operating instructions for movement of the table 110. The system 100 may further include an energy source 114 configured to provide energy to the powder on the substrate 106. Energy (e.g., electromagnetic energy) from the energy source 114 may be directed to the substrate 106 and the structure 108 through a mirror 116, which may orient the energy to the substrate 106. The energy source 114 may comprise, for example, a laser (e.g., selective laser additive manufacturing), an electron beam, a source of microwave energy, or another energy source. In some embodiments, powder from the powder delivery nozzle 104 is disposed on the substrate 106 and simultaneously exposed to energy (illustrated by broken lines 118) from the energy source 114. Although FIG. 1 illustrates that the table 110 is operably coupled with the central processing unit 112 to effect movement of table 110, the disclosure is not so limited. In other embodiments, the central processing unit 112 is operably coupled with the powder delivery nozzle 104 and the energy source 114 and the powder delivery nozzle 104 and the energy source 114 is configured to move in one or more directions (e.g., the x-direction, the y-direction, and the z-direction) responsive to receipt of instructions from the central processing unit 112. In some such embodiments, one or some of the powder delivery nozzle 104, the energy source 114 and the table 110 may be configured to move in one or more directions. Movement of the powder delivery nozzle 104, the energy source 114, the table 110, or both may facilitate forming the structure 108 to have a desired composition and geometry. In use and operation, a layer of powder from the powder feed 102 and expelled by the powder delivery nozzle 104 may be formed over the substrate 106 and subsequently exposed to energy from the energy source 114 to form inter-granular bonds between particles of the layer of powder. In other embodiments, the powder is exposed to energy from the energy source 114 substantially simultaneously with delivery of the powder to the surface of the substrate 106 or substantially immediately thereafter. In some such embodiments, portions of the layer of the structure 108 being formed may be exposed to energy from the energy source 114 prior to formation of the entire layer of the structure 108. At least one of the energy source 114 and the table 110 may be configured to move responsive to instructions from the central processing unit 112. After formation of the layer of the structure 108, the substrate 106 is moved away from the energy source 114, such as by movement of one or both of the table 110 and the energy source 114 responsive to receipt of instructions from the central processing unit 112. Additional powder may be delivered to the surface of the previously formed layer of the structure 108 in a desired pattern and exposed to energy from the energy source 114 to form inter-granular bonds between adjacent particles of the powder in the layer and between particles of the powder in the layer and the underlying layer of the structure 108. Each layer of the structure 108 may be between about 25 μm (about 0.001 inch) and about 500 μm (about 0.020 inch), such as between about 25 μm and about 50 μm, between about 50 μm and about 100 μm, between about 100 μm and about 200 μm, between about 200 μm and about 300 μm, between about 300 μm and about 400 μm, or between about 400 μm and about 500 μm. Accordingly, the structure 108 may be formed one layer at a time, each layer having a thickness between about 25 μm and about 500 μm. In some embodiments, one or more layers of the structure 108 may be formed to exhibit a different composition than one or more other layers of the structure 108. In some embodiments, different portions of a single layer of the structure 108 may exhibit a different composition than other portions of the same layer of the structure 108. By way of nonlimiting example, where the structure 108 comprises a fuel rod, a portion (i.e., a central portion) of the layer may comprise a nuclear fuel (e.g., uranium oxide) and a portion (i.e., a peripheral portion) of the layer may comprise a cladding material. Where the structure 108 comprises a portion of a reactor core, portions of the layer may comprise a fuel rod and other portions of the layer may comprise a reactor poison. In some embodiments, the structure 108, or at least a portion thereof, may be formed to exhibit a void fraction (e.g., porosity) between about 0.10 and about 0.50, such as between about 0.10 and about 0.20, between about 0.20 and about 0.30, between about 0.30 and about 0.40, or between about 0.40 and about 0.50. In some such embodiments, the structure 108 may not be formed to a full theoretical density thereof. As will be described herein, forming the structure 108 to exhibit a void fraction may facilitate expansion of the structure 108 during use thereof in a nuclear reactor. By way of nonlimiting example, a nuclear fuel may be formed to exhibit a void fraction according to the methods described herein. In some embodiments, a nuclear fuel exhibiting a void fraction between about 0.10 and about 0.50, as described above, may be formed by additive manufacturing. The nuclear fuel may be used in a fuel rod in, for example, a fast-neutron reactor, a light water reactor, a modular nuclear reactor, a space reactor, a micro reactor, or other nuclear reactor. FIG. 2A and FIG. 2B are a respective simplified perspective view and a cross-sectional view of a fuel rod 200 formed according to the methods described herein, in accordance with embodiments of the disclosure. The fuel rod 200 includes a nuclear fuel 202 surrounded by cladding 204. A fission barrier material 206 may be disposed between the nuclear fuel 202 and the cladding 204. In some embodiments, the fission barrier material 206 may substantially conformally overlie the nuclear fuel material 202. In some such embodiments, the fission barrier material 206 may have substantially the same cross-sectional shape as the nuclear fuel material 202. Similarly, the cladding 204 may substantially conformally overlie the fission barrier material 206 and may have substantially the same cross-sectional shape as the nuclear fuel material 202 and the fission barrier material 206. The nuclear fuel 202 may include any suitable nuclear fuel. In some embodiments, the nuclear fuel 202 comprises uranium dispersed in zirconium (e.g., U-10Zr (an alloy of uranium and about 10 weight percent zirconium)). The nuclear fuel 202 may comprise a metallic foam and may exhibit a void fraction between about 0.10 and about 0.50, as described above. In some embodiments, the void fraction of the nuclear fuel 202 may be greater than about 0.10, greater than about 0.20, greater than about 0.30, greater than about 0.40, or even greater than about 0.50. In some such embodiments, the nuclear fuel 202 may exhibit a theoretical density less than about 100%, such as less than about 95%, less than about 90%, less than about 85%, less than about 80%, or even less than about 70%. In some embodiments, a composition of the nuclear fuel 202 varies with a distance (e.g., a radial distance) from a center of the fuel rod 200. By way of nonlimiting example, an enrichment of the nuclear fuel 202 may increase with an increasing distance from a center of the nuclear fuel 202 and may decrease with a distance from the cladding 204. In other words, the enrichment of the nuclear fuel 202 may be greater proximate the cladding 204 than proximate a center of the nuclear fuel 202. In other embodiments, the enrichment of the nuclear fuel 202 may be less proximate the cladding 204 than proximate the center of the nuclear fuel 202. Forming the nuclear fuel 202 by additive manufacturing may facilitate forming the nuclear fuel 202 to exhibit a varying amount of enrichment with varying distance from a center thereof. In some embodiments, a composition of the nuclear fuel 202 may vary along a length of the fuel rod 200. By way of nonlimiting example, the nuclear fuel 202 may exhibit a greater amount of enrichment at a top and bottom of the fuel rod 200 than at a longitudinally central portion of the fuel rod 200. Forming the nuclear fuel 202 by additive manufacturing and layer by layer may facilitate forming the nuclear fuel 202 to exhibit a varying enrichment along a longitudinal axis thereof. In some embodiments, the nuclear fuel 202 may include one or more dopants, such as one or more of a metal oxide (aluminum oxide, zirconium oxide, etc.), carbon nanotubes, carbon nanotubes coated with a metal oxide (e.g., aluminum oxide, zirconium oxide, another metal oxide, or combinations thereof), another material, or combinations thereof. In some embodiments, the dopant may facilitate improved tensile strength of the nuclear fuel 202. The nuclear fuel 202 including the one or more dopants therein may be formed by additive manufacturing with a powder comprising the nuclear fuel and the one or more dopants dispersed therein. The cladding 204 may comprise stainless steel (e.g., austenitic 304 stainless steel, 316 stainless steel, HT-9 stainless steel (a ferritic steel comprising about 12.3 weight percent chromium, about 0.5 weight percent nickel, about 1.0 weight percent molybdenum, about 0.01 weight percent copper, about 0.3 weight percent vanadium, about 0.47 weight percent vanadium, the remainder comprising carbon, manganese, phosphorus), stainless steels including alloys of chromium and nickel), an oxide dispersion-strengthened alloy (ODS) including one or more nickel-based alloys, iron-based alloys, and aluminum-based alloys such as, for example, iron aluminide, iron chromium, iron-chromium-aluminum, nickel chromium, and nickel aluminide, a nano-ferritic alloy (NFA), a zirconium-based alloy, another material, or combinations thereof. A thickness of the cladding 204 may be between about 0.5 μm and about 800 μm, such as between about 0.5 μm and 1.0 μm, between about 1.0 μm and about 5.0 μm, between about 5.0 μm and about 25 μm, between about 25 μm and about 50 μm, between about 50 μm and about 100 μm, between about 100 μm and about 250 μm, between about 250 μm and about 500 μm, or between about 500 μm and about 800 μm. Forming the fuel rod 200 to comprise the cladding 204 and the nuclear fuel 202 concurrently may reduce thermal contact resistance between the cladding 204 and the nuclear fuel 202. The fission barrier material 206 may comprise zirconium, vanadium, another material, and combinations thereof. In some embodiments, the fission barrier material 206 is substantially free of pinholes such that fission products do not substantially diffuse from the nuclear fuel material 202 through the fission barrier material 206 and to the cladding 204. In some such embodiments, the fission barrier material 206 is hermetically disposed around the nuclear fuel material 202. Accordingly, the fission barrier material 206 may impede or reduce so-called fuel-cladding mechanical and chemical interactions (FCCI). By way of contrast, conventional fission barrier materials formed by atomic layer deposition or chemical vapor deposition and may include holes or deformities through which fission products may propagate. Alternatively, fission barrier products may be formed with a foil that lines the fuel material, however the foil may unwrap during fabrication of the fuel element associated with the foil. Without wishing to be bound by any particular theory, it is believed that forming the fission barrier material 206 around the nuclear fuel material 202 by additive manufacturing may facilitate a fission barrier material hermetically disposed around the nuclear fuel material 202 and may reduce a likelihood of diffusion of fission products through the fission barrier material 206. A thickness of the fission barrier material 206, exaggerated in the view of FIG. 2B for clarity, may be between about 0.5 μm and about 500 μm, such as between about 0.5 μm and about 5.0 μm, between about 5.0 μm and about 25 μm, between about 25 μm and about 50 μm, between about 50 μm and about 100 μm, between about 100 μm and about 250 μm, or between about 250 μm and about 500 μm. In some embodiments, an interface between the fission barrier material 206 and each of the nuclear fuel 202 and the cladding 204 may comprise a gradient. In some such embodiments, an atomic percent of components of the fission barrier material 206 (e.g., vanadium, zirconium, etc.) may increase from a location proximate the nuclear fuel material 202 to a location at a radially central portion of the fission barrier material 206. An atomic percent of the components of the fission barrier material 206 may decrease from a location proximate the radially central portion of the fission barrier material 206 to a location proximate the cladding 204. In some embodiments, the atomic percent of the components of the fission barrier material 206 may not be uniform and may vary with a distance from the center of the fuel rod 200. Referring to FIG. 2B, the fuel rod 200 may include lobes 208 and corresponding valleys 210 between adjacent lobes 208. The lobes 208 may protrude further from a center of the fuel rod 200 than the valleys 210. The lobes 208 and the valleys 210 may increase an exposed surface area of the fuel rod 200 for a given cross-sectional area of the fuel rod 200 and may improve heat transfer between the fuel rod 200 and fluids surrounding the fuel rod 200 in use and operation of a nuclear reactor including the fuel rod 200. Accordingly, along a length of the fuel rod 208, 200 the valleys 210 and lobes 208 may define a volume through which a cooling fluid (e.g., air) may flow during use and operation of the fuel rod 200 in a nuclear reactor. Referring to FIG. 2A, the fuel rod 200 may twist, for example in a helical configuration, along a longitudinal axis thereof. In other words, along the longitudinal axis of the fuel rod 200, the locations of the lobes 208 and the valleys 210 may rotate. In some such embodiments, the fuel rod 200 may include a helical gaseous heat exchange structure. In some such embodiments, the fuel rod 200 may be additively manufactured by forming at least one layer of the fuel rod 200 on an adjacent layer, the at least one layer rotated (with respect to a longitudinal axis) relative to the adjacent layer. In other words, the at least one layer may, for example, have a same cross-sectional shape as the adjacent layer, but may be rotated (e.g., by about 1°, by about 2°, by about 5°, by about 10°, by about 15°) with respect to the adjacent layer. In some such embodiments, the at least one layer may be said to be rotationally offset with respect to the adjacent layer. Although FIG. 2B illustrates that the fuel rod 200 includes four lobes 208, the disclosure is not so limited. In other embodiments, the fuel rod 200 includes fewer lobes 208 (e.g., three lobes) or a greater number of lobes 208 (e.g., five lobes, six lobes, seven lobes, eight lobes, etc.). A distance D between an exterior surface of a lobe 208 to an exterior surface of an opposing lobe 208 may be between about 0.25 cm to about 1.0 cm, such as between about 0.25 cm and about 0.35 cm, between about 0.35 cm and about 0.5 cm, between about 0.5 cm and about 0.75 cm, or between about 0.75 cm and about 1.0 cm. In some embodiments, the distance D is equal to about 0.635 cm (about 0.25 inch). In some embodiments, the fuel rod 200 may include one or more neutron reflector materials. By way of nonlimiting example, the fuel rod 200 may include a neutron reflector material at a top thereof, at a bottom thereof, or both. The neutron reflector material may include beryllium, graphite, another material, or combinations thereof. In some embodiments, the neutron reflector material may be formed by the additive manufacturing process. By way of nonlimiting example, a layer of the neutron reflector material may be formed by additive manufacturing, layers of the nuclear fuel 202 and cladding 204 may be formed by additive manufacturing, and additional layers of the neutron reflector material may be formed by additive manufacturing over the layers of the nuclear fuel 202 and the cladding 204. In some embodiments, the nuclear fuel 202 includes one or more grids, one or more tetrahedra, or a combination thereof within the nuclear fuel 202 to improve mechanical stability of the nuclear fuel 202 during irradiation and fission. FIG. 2C is a perspective view of nuclear fuel 202′ including support structures 230 having a tetrahedral shape. The support structures 230 may comprise any of the nuclear fuels described above, such as, for example, uranium, uranium oxide, uranium dioxide, uranium nitrides, uranium borides, a transuranic material, thorium, oxides, thereof, or another nuclear fuel. FIG. 2D is a perspective view of a nuclear fuel 202″, in accordance with other embodiments of the disclosure. The nuclear fuel 202″ includes support structures 230′ extending throughout a body of the nuclear fuel 202″. The support structures 230′ may include a nuclear fuel, as described above with reference to the support structure 230 (FIG. 2C). In some embodiments, spaces between the support structures 230, 230′ may be void and the nuclear fuel 202′, 202″ may comprise a porous nuclear fuel. In addition, compartmentalization of the fissile materials with high temperature refractory metal oxides such as zirconium oxide and refractory metals such as molybdenum, tungsten, chromium, vanadium, tantalum, rhenium, hafnium, titanium, or other refractory metal may further impede creep (e.g., slump) during high temperature excursions, power excursions, or both. In other words, forming the nuclear fuels 202, 202′, 202″ to include high temperature refractory metal oxides and refractor metals may improve operating properties of corresponding fuel rods, such as by reducing or impeding creep during high temperature excursions, power excursions, or both. FIG. 3 is a simplified perspective view of multiple fuel rods 200 arranged in a bundle 300, in accordance with embodiments of the disclosure. As illustrated in FIG. 3, the shape of the fuel rods 200 may facilitate an increased packing density of the fuel rods 200 in the bundle 300. In some embodiments, a lobe 208 (FIG. 2B) of one fuel rod 200 may be disposed in a valley 210 (FIG. 2B) of an adjacent fuel rod 200. FIG. 4 is a cross-sectional view of a bundle 400 of fuel rods 401, in accordance with other embodiments of the disclosure. The fuel rods 401 include a nuclear fuel material 402 surrounded by a cladding material 404. In some embodiments, a fission barrier material 406 may intervene between the nuclear fuel material 402 and the cladding material 404, as described above with reference to the fuel rod 200 (FIG. 2A and FIG. 2B). The fuel rods 401 may have a cylindrical shape with a circular cross-sectional shape. The nuclear fuel material 402 may have a same cross-sectional shape as described above with reference to the nuclear fuel material 202 (FIG. 2A and FIG. 2B). The fission barrier material 406 may be conformally disposed around a periphery of the nuclear fuel material 402. In other embodiments, the fission barrier material 406 may be disposed around an inner surface of the cladding 404. The fission barrier material 406 may reduce or prevent chemical interactions between the nuclear fuel material 402 and the cladding 404 and may reduce or prevent stress corrosion cracking of the cladding 404. The cladding 404 may be disposed around the fission barrier material 406. The cladding 404 may have a circular outer circumference. Accordingly, the fuel rod 401 may include a nuclear fuel material 402 having a different cross-sectional shape than the cross-sectional shape of the cladding 404. FIG. 5 is a simplified cross-sectional view of a fuel rod 500, in accordance with embodiments of the disclosure. The fuel rod 500 includes a nuclear fuel material 502 and cladding 504 surrounding the nuclear fuel material 502. The nuclear fuel material 502 and the cladding 504 may include the same materials described above with reference to the nuclear fuel material 202 and the cladding 204. In some embodiments, the fuel rod 500 includes a fission barrier material 506, which may include the same materials described above with reference to the fission barrier material 206. The fuel rod 500 may include an aperture 520 extending through a central axis thereof. The aperture 520 may be defined by a circular cross-sectional shape. The aperture 520 may extend along a length of the fuel rod 500. In some embodiments, a heat transfer fluid (e.g., water, a vapor, etc.) may be flowed through the aperture 520 during use and operation of a nuclear reactor including the fuel rod 500 to recover heat from the fuel rod 500. In some embodiments, the aperture 520 may be configured to receive a heat pipe for recovering thermal energy from the fuel rod 500 and transferring the thermal energy to another portion of a nuclear reactor associated with the fuel rod. In some embodiments, a heat pipe may be formed by additive manufacturing concurrently with forming the fuel rod 500. Although FIG. 5 illustrates that the aperture 520 has a circular cross-sectional shape, the disclosure is not so limited. In other embodiments, the aperture 520 has a different cross-sectional shape, such as elliptical, oval, square, rectangular, triangular, hexagonal, or another shape. Although FIG. 5 illustrates that the cladding 504 has the same cross-sectional shape as the nuclear fuel material 502, the disclosure is not so limited. In other embodiments, an outer surface of the cladding 504 may have a circular cross-sectional shape, as described above with reference to the cladding 404 (FIG. 4). FIG. 6 is a simplified cross-sectional view of a fuel rod 600, in accordance with embodiments of the disclosure. The fuel rod 600 may be substantially the same as the fuel rod 500 (FIG. 5), except that the fuel rod 600 may include multiple apertures 620 extending therethrough. The apertures 620 may define flow channels sized and shaped to receive a heat transfer fluid, such as water or air, during use and operation of the fuel rod 600. In use and operation, the heat transfer fluid may be flowed through the apertures 620 for transferring thermal energy from the nuclear fuel material 602 to the heat transfer fluid and to other portions of a nuclear reactor core associated with the nuclear reactor core. In some embodiments, a nuclear reactor core may comprise a bundle of fuel rods, such as the fuel rods 200, 500, 600 described above. In some embodiments, a poison rod comprising a burnable poison may be disposed in the bundle at desired axial and radial locations. Accordingly, a bundle may comprise fuel rods and at least one poison rod. The poison rods may include burnable poison materials, such as boron, gadolinium, Gd2O3, B4C, etc., another material exhibiting a high thermal neutron absorption cross-section, and combinations thereof. The poison rods may be formed by additive manufacturing processes. In some such embodiments, the nuclear reactor core may be formed by additive manufacturing. By way of nonlimiting example, layers of the nuclear reactor core, at least some of the layers (e.g., each layer) comprising materials of the fuel rod (e.g., the nuclear fuel, cladding, and the diffusion barrier material) and materials of the burnable poison may be formed one over the other by additive manufacturing until a nuclear reactor core having a desired size and shape is formed. Similarly, burnable poisons may be layered over individual particles of fuel embedded within the fuel matrix of the nuclear fuels described above. In some such embodiments, the powder feed 102 (FIG. 1) may comprise at least some particles comprising the nuclear fuel material coated with a layer of a burnable poison. By way of nonlimiting example, the burnable poison may be formed over the particles of the fuel material by atomic layer deposition (ALD), chemical vapor deposition (CVD), physical vapor deposition (PVD), another deposition method, or combinations thereof. In some embodiments, a nuclear reactor core may be assembled from individual plates produced using additive manufacturing techniques. Referring to FIG. 7, an assembly 700 comprising heat plates 702 stacked in an arrangement that may comprise a portion of a reactor core. The heat plates 702 may comprise a suitable material, such as stainless steel (e.g., 304 stainless steel), zirconium, or another material. The plates may include mechanical interfaces for a heat exchange structure 704, such as a heat exchangers, a heat pipes, or a combination thereof. The heat exchange structures 704 may extend through the heat plates 702. The heat exchange structure 704 may extend from the reactor core proximate a nuclear fuel material to another portion of a nuclear reactor and may be configured for transferring heat from the nuclear fuel to the another portion of a nuclear reactor. In some embodiments, the heat plates 702 are substantially the same and comprise substantially the same geometry. In other words, the heat exchange structure 704 may extend through each heat plate 702 at substantially the same locations on the respective heat plate 702. Stacking of the individual plates in a critical assembly allows a reactor core to be constructed. Although not illustrated in FIG. 7, fuel rods (e.g., the fuel rods 200, 401, 500, 600) described above with reference to FIG. 2A through FIG. 6 may extend through one or more of the heat plates 702. In some embodiments, the assembly 700 may be formed in an additive manufacturing process such that the assembly 700 is formed and arranged (e.g., the heat plates 702 are stacked with the heat exchange structure 704 and fuel rods extending therethrough) during formation of each heat plate 702. In yet other embodiments, the heat plates 702 may be formed by additive manufacturing to include mechanical interfaces for coupling the heat exchange structures 704 and the fuel rods thereto. Each of the fuel rods 200, 401, 500, 600, the bundles 300, 400, and the assembly 700 may be formed by additive manufacturing processes with the system 100 (FIG. 1) described above. In some such embodiments, the fuel rods, bundles, and assembly are formed layer by layer be disposing a layer of a powder material on a surface of a substrate or a previously-formed layer, exposing the powder material to energy from the energy source, forming another layer of a powder material on the surface of the previously-formed layer, exposing the another layer of the powder material to energy, and repeating the disposing powder material and exposing the powder material to energy until a structure having a desired size and shape is formed. As described above, exposing the powder material to energy from the energy source may form inter-granular bonds between the particles of the powder material and underlying layers of the structure being formed. In some embodiments, at least a portion of the structure being formed may not be formed to a substantially full theoretical density after the additive manufacturing process. In some such embodiments, the structure formed by the additive manufacturing process may be exposed to annealing conditions to densify at least a portion of the article. In some embodiments, after the structure 108 is formed by additive manufacturing with the system 100, the structure 108 may be densified. Densifying the structure 108 may be performed by swaging, extrusion, hot isostatic pressing, thermal soaking, laser annealing, or combinations thereof. In some embodiments, the structure 108 is exposed to annealing (e.g., sintering) conditions to densify at least a portion thereof. In some embodiments, the structure 108 may be subjected to a hot isostatic pressing (HIP) process, a sintering process, a spark plasma sintering (SPS) process, or other densification process. By way of nonlimiting example, the structure 108 may be sintered to substantially fully densify the structure thereof. In some embodiments, densifying the structure 108 may include densifying at least a portion of the structure 108 while substantially not densifying at least another portion of the structure 108. By way of nonlimiting example, with reference to FIG. 2A and FIG. 2B, the cladding 204 may be annealed and densified, while the nuclear fuel material 202 is not annealed. In some such embodiments, exposing the structure 108 to annealing conditions may include exposing the structure 108 to a melting temperature greater than a melting temperature of the cladding 204 and lower than a melting temperature of the nuclear fuel material 202. Without wishing to be bound by any particular theory, it is believed that because the nuclear fuel material 202 exhibits a higher melting temperature than the cladding 204, the nuclear fuel material 202 is not densified or annealed while the cladding 204 is annealed. As another nonlimiting example and with reference to FIG. 7, the heat plates 702 and heat exchange structures 704 may be annealed and densified while nuclear fuel materials in fuel rods associated with the assembly 700 are not substantially densified. Exposing the structure 108 to annealing conditions may include disposing the structure 108 in a die and exposing the structure 108 to a suitable temperature and pressure to densify at least a portion of the structure 108. In some embodiments, the die may not comprise graphite to prevent formation of binary or ternary compounds comprising carbon. In some such embodiments, the die may comprise tungsten carbide, molybdenum, or a combination thereof. Where the die comprises graphite, the surfaces of the sintered structure 108 may be exposed to a reducing agent to remove any carbon contamination (e.g., binary or ternary compounds including carbon) from the sintered structure 108, such as by exposing the surfaces of the sintered structure 108 to hydrogen gas. FIG. 8 is a simplified flow diagram illustrating a method 800 of forming an article, in accordance with embodiments of the disclosure. The method 800 includes act 802 including forming one or more powder mixtures; act 804 including disposing a powder material on a surface of a substrate; act 806 including exposing the powder material to energy from an energy source to form inter-granular bonds between particles of the powder material and form a layer of an article; act 808 including repeating act 802 through 806 until an article having a desired size, shape, and composition is formed; and act 810 including exposing the article to annealing conditions to densify at least a portion of the article. Act 802 includes forming one or more powder mixtures. Forming the one or more powder mixtures may include forming a powder mixture for each component of an article (e.g., a nuclear fuel material, a fuel rod, a bundle of fuel rods, etc.) to be formed. By way of nonlimiting example, a powder mixture may be formed to include particles of a nuclear fuel material, such as particles of uranium and a metal material such as at least one of zirconium and tungsten. In some embodiments, the uranium and the metal material may not be mixed. Other powder materials may include particles of a cladding material, particles of a diffusion barrier material, particles of a poison material, particles of a heat pipe, one or more dopant materials to be interspersed in a nuclear fuel material, and combinations thereof. Act 804 includes disposing a powder material on a surface of a substrate. The substrate may comprise a previously-formed layer of the article. The powder material may be disposed on the surface of the substrate such that the layer of the article being formed exhibits a desired cross-sectional composition. In some embodiments, particles of cladding material may be disposed around particles of a nuclear fuel material. Particles of a fission barrier material may be disposed between the particles of the nuclear fuel material and the particles of the cladding material. In some embodiments, the powder material may further include particles of a neutron reflector material, particles of a burnable poison material, particles of a heat pipe, or particles of another material. In some embodiments, the substrate may be heated while disposing the powder material on a surface of the substrate. By way of nonlimiting example, the substrate may be heated to between about 50° C. and about 150° C., such as between about 50° C. and about 100° C. or between about 100° C. and about 150° C. Act 806 includes exposing the powder material to energy from an energy source to form inter-granular bonds between particles of the powder material and form a layer of an article. In some embodiments act 804 and act 806 occur substantially at the same time. In other words, in some such embodiments, the powder material may be exposed to energy from the energy source substantially simultaneously with disposing the powder material on the surface of the substrate. Act 808 includes repeating act 802 through 808 until an article having a desired size, shape, and composition is formed. In some embodiments, one or more layers of the article may be formed to exhibit a different composition than one or more other layers of the article. By way of nonlimiting example, in some embodiments, a layer comprising a nuclear fuel material and a cladding material may be formed over one or more layers comprising a neutron reflector material surrounded by a cladding material. After forming the nuclear fuel and the cladding to a desired dimension, one or more layers of the neutron reflector material surrounded by the cladding material may be formed over the nuclear fuel and the cladding to form a fuel rod comprising the nuclear fuel with a neutron reflector above and below the nuclear fuel, the neutron reflectors and the nuclear fuel surrounded by the cladding. Act 810 includes exposing the article to annealing conditions to densify at least a portion of the article. In some embodiments, the article is removed from the additive manufacturing tool and disposed in a die to densify at least a portion of the article. Exposing the article to annealing conditions may include densifying, for example, the cladding while substantially not densifying the nuclear fuel material. In some such embodiments, the article is exposed to a temperature between about 700° C. and about 1,600° C., such as between about 700° C. and about 800° C., between about 800° C. and about 1,000° C., between about 1,000° C. and about 1,200° C., between about 1,200° C. and about 1,400° C., or between about 1,400° C. and about 1,600° C. Forming the nuclear fuel to comprise a metallic foam and exhibit a porosity as described herein may facilitate operation of a fuel cell including the nuclear fuel without the use of sodium or other another material (e.g., helium) between the nuclear fuel and the cladding. The porosity of the nuclear fuel may facilitate reduced expansion of the nuclear fuel during operation of the nuclear reactor including the fuel rod and may reduce undesired stresses on the cladding wall. In other words, the porosity of the nuclear fuel incorporates sufficient void space to accommodate the swelling of the nuclear fuel without causing stress on the cladding. In addition, forming the fuel rod without sodium may eliminate sodium bonding between sodium and the cladding and eliminate requirements of reprocessing of sodium after use of the fuel rod. The nuclear fuel may be formed to include one or more dopants therein to reduce or prevent melting or sintering thereof during operating conditions of a fuel rod including the nuclear fuel. In some embodiments, the one or more dopants may reduce or prevent creep (i.e., slumping) of the fuel rod during operating conditions thereof when the fuel rod is exposed to excessive temperatures. Accordingly, one or more components of a nuclear reactor (e.g., a fuel rod) may be fabricated with internal features, external features, or both for coupling the component to heat exchange mechanisms or power conversion mechanisms. In some embodiments, a fuel rod includes one or more annular spaces within the nuclear fuel material through which a cooling fluid may flow. Forming structures according to embodiments described herein may facilitate fabrication of intricately coupled fuel cladding systems, rapid production of components for a nuclear reactor, implementation of complex physical features of structures of a nuclear reactor, reduction in fabrication time and material waste, and formation of gradients in components (e.g., a nuclear fuel) of the nuclear reactor. In addition, one or more components of a nuclear reactor may be formed without welds connecting different portions of a nuclear reactor core. While embodiments of the disclosure may be susceptible to various modifications and alternative forms, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the disclosure is not limited to the particular forms disclosed. Rather, the disclosure encompasses all modifications, variations, combinations, and alternatives falling within the scope of the disclosure as defined by the following appended claims and their legal equivalents. |
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description | A method for evaluating tool performance is implemented by a die evaluation software program installed on a computer such as a Pentium II(copyright) running Windows 95(copyright) and having user interface (not shown), for example, a keyboard, a cathode ray tube (CRT) monitor and a mouse. Alternatively, the program could be adapted to run under LINUX or Mac OS(copyright) or other operating system. In one embodiment the die evaluation program runs under a spreadsheet program having a macro creation feature, for example, Microsoft Excel version 7 running under Microsoft Office 95. The spreadsheet program is loaded onto the computer and the die evaluation program is loaded under the spreadsheet program and saved as an Excel application program. A user enters identifying and historical information pertaining to a die and associated punches onto spreadsheets (see FIGS. 1 through 6) displayable on the CRT and which are maintained in a computer database (not shown) over time. After the user has entered initial identifying and historic data for a particular die and punches, the user enters subsequent servicing data into the database for the die and punches at times when servicing is performed. FIG. 1 illustrates a New Die Data Entry Sheet 10 for entering initial identifying and historical information for a die and punches into the computer. Each new die is assigned a die identification number 12 when entered into the computer. The user enters die-related information into New Die Data Entry Sheet 10 by filling in, for example, a plant location cell 14, rotor and stator number cells 16, a die manufacturer cell 18, a die dimension cell 20, a service date cell 22 (i.e. date on which die was put into service), a tonnage cell 24, a strip width and thickness cell 26, a progression cell 28, a total die weight cell 30, an exit method cell 32, a die serial number cell 34, an amount punches enter die cell 36, a number of rows in die cell 38, and a die weekly line rate cell 40. Automated information entry, such as automatic gauging, bar code scanning or OCR scanning of forms can replace some of this data entry for greater speed and accuracy, if desired. Station number cells 42 identify each punch and die station in the die to be tracked by the die evaluation program. New Die Data Entry Sheet 10 includes data for each punch in a punch and die station, for example, a punch description cell 44, a punch height cell 46, a cell 48 for stripper plate thickness minus any counter-bores, and a die height cell 50. New Die Data Entry Sheet 10 also includes, for example, rotor/stator strike plate thickness cells 52, a cell 54 to indicate whether 0.090-inch punch safety margins are desired, a cell 56 for laminations per stack, cells 58 and 60 for goal amounts respectively for punch and die grinds, a cell 62 for goal number of hits per grind, a cell 64 for goal number of hits per die set, and cells 66 for other historical information predating entry of the die into the die evaluation program. When New Die Data Entry Sheet 10 has been completed, the user activates (for example, via mouse) a Create New Worksheets Macro Button 68, to create additional spreadsheets (not shown in FIG. 1) for the newly entered die. One of the additional spreadsheets, a Die Summary Sheet (not shown in FIG. 1), can be displayed by activating a Return To Summary macro button 76. The die evaluation program automatically transfers New Die Data Entry Sheet 10 information into corresponding cells in the additional spreadsheets. For example, FIG. 2 illustrates a Die Life Data Sheet 70 displaying some of the cell data originally entered by the user into New Die Data Entry Sheet 10. Die Life Data Sheet 70 automatically includes die identification number 12, plant location cell 14, rotor and stator number cells 16, die manufacturer cell 18, die dimension cell 20, service date cell 22 (i.e. date on which die was put into service), tonnage cell 24, strip width and thickness cell 26, progression cell 28, total die weight cell 30, exit method cell 32, die serial number cell 34, and amount punches enter die cell 36. Die Life Data Sheet 70 displays data for each punch and die station in the newly entered die, for example, station number cells 42, punch description or punch and die description cells 44, punch height cells 46, stripper plate thickness minus any counter-bores cells 48 and die height cells 50. Die Life Data Sheet 70 also includes rotor/stator strike plate thickness cells 52 and cell 54 indicating whether 0.090-inch punch safety margins are desired. The user inserts into Die Life Data Sheet 70 a standard height dimension for non-usable die life 72. A cell 74, to check for punch rebuild or repair, provides a warning message xe2x80x9cNOWxe2x80x9d (not shown) when 0.090-inch safety margin cell 54 has been activated and it is determined that a punch for a corresponding station has reached a value within 0.090 inches of bottoming out on a corresponding stripper plate. When 0.090-inch safety margin cell 54 has not been activated, check for punch rebuild or repair cell 74 provides a warning xe2x80x9cLESS THAN 0.030!xe2x80x9d (not shown) when it is determined that a punch for a corresponding station has reached a value within 0.030 inches of bottoming out on a corresponding stripper plate. If rotor/stator strike plate thickness cells 52 indicate that spring loaded interlock strike plate thickness for. a die has decreased to less than 0.125 inches, check for punch rebuild or repair cell 74 also provides a xe2x80x9cNOWxe2x80x9d warning (not shown). Die Life Data Sheet 70 also is automatically updated with information entered by the user, for example, into a Die Service Record 80 as shown in FIG. 3. Die Service Record 80 receives information cumulatively entered by the user to describe punch or die servicing or repair. When a die is serviced or repaired the user completes, for example, a cell 82 for current service date, a cell 84 for total number of hits on a just-completed press run, a cell 86 for amount ground from die during present servicing, a cell 88 for amount ground from punch during present servicing, a cell 90 for reason for servicing, a cell 92 for repair parts used, a cell 94 for person performing servicing, and a comments cell 96. Die Service Record 80 displays other information automatically determined by the die evaluation program when servicing information is entered onto Die Service Record 80 for a particular die. For example, the program generates a histogram 98 of frequency for reason die was sent to repair. The program also determines, for example, hits since last grind 100, whether die needs full grind 102, total hits on die to date 104, total amount ground from die to date 106, remaining die height 108, total amount ground from punch to date 110, and remaining punch height 112. Die Service Record 80 also displays short-term die accumulation data 114, historical data from prior to Die Service Record 116, and long-term die accumulation data 118 (a combination of short-term die accumulation data 114 with historical data from prior to Die Service Record 116). When, for example, Die Service Record 80 has been filled, the user can activate Reset Service Record macro button 124 to make room for entering additional servicing information and to update long-term die accumulation data cells 118. The die evaluation program uses information from Die Service Record 80 to update Die Life Data Sheet 70 (shown in FIG. 2). For example, Die Life. Data Sheet 70 cells for total amount ground from punch to date 110 and total amount ground from die to date 106 are automatically updated each time Die Service Record 80 is updated. Cells for remaining usable punch life 120 and remaining usable die life 122 also are determined from Die Service Record 80 data. Cell 120 for remaining usable punch life is linked to 0.090 safety margin cell 54 and will display values including safety margins if the user has activated cell 54. Cell 122 for remaining usable die life is linked to cell 72 for standard height dimension for non-usable die life and will display values determined by the die evaluation program using cell 72 value. The die evaluation program uses information in Die Service Record 80, Die Life Data Sheet 70 and New Die Data Entry Sheet 10 (shown in FIG. 1) to determine a die useful life in terms of weeks remaining. For example, FIG. 4 illustrates a Weeks Remaining Sheet 128 including data determined from Die Service Record 80 and data transferred from New Die Data Entry Sheet 10 and Die Life Data Sheet 70. Weeks Remaining Sheet 128 includes data for each punch and die station in the die identified by identification number 12, for example, punch and die description 44, punch life remaining 120, and die life remaining 122. The die evaluation program determines cells 132 for punch grinds remaining, cells 134 for die grinds remaining, cells 136 for cores or laminations remaining before re-punch, cells 138 for cores or laminations remaining before die life is exhausted, cells 140 for weeks remaining before re-punch, and cells 142 for weeks remaining before die life is exhausted. The die evaluation program makes the foregoing determinations based on values in cells 144 for average amount of punch grind, cells 146 for average amount of die grind, cells 148 for average number of hits per grind, and cells 150 for number of laminations. Weeks Remaining Sheet 128 displays a safety margin indicator 130 indicating whether safety margin cell 54 has been activated from Die Life Data Sheet 70 (shown in FIG. 2). Values in cells 144, 146, 148 and 150 are entered by the user. The program also determines values for cell 152 for average number of weeks remaining before re-punch, cell 154 for average number of weeks remaining before die life is exhausted, cell 156 for minimum number of weeks remaining before re-punch and cell 158 for minimum number of weeks remaining before die life is exhausted. Weeks Remaining Sheet 128 also displays New Die Data Entry Sheet 10 cells 38 for die number of rows and 40 for current weekly line rate. FIG. 5 illustrates a Die Performance Sheet 170 including sections 212, 174, 176, 178 and 196 for die demographics, die measurements, benchmark data, performance results, and efficiency ratings. The user activates cell 210 to select whether information on Die Performance Sheet 170 is to be determined with short-term or long-term data maintained in Die Service Record 80. FIG. 5 depicts Die Performance Sheet 170 with long-term data selected, and FIG. 6 depicts Die Performance Sheet 170 with short-term data selected. Section 212 includes cells for plant location 14, date of most recent service 82 (from Die Service Record 80), and time period to date 172. If the user selects short-term data via cell 210, time period to date 172 is determined as a time period between earliest service date 82 shown on Die Service Record 80 and most recent service date 82. If long-term data is selected, time period to date 172 is determined as a time period between date die was put into service 22 and most recent service date 82. Die and punch measurements 174 over time period to date 172 are automatically entered into Die Performance Sheet 170 from Die Service Record 80 and include, for example, a total number of hits to date cell 214. Benchmark data 176 include a cell 160 for goal hits per grind, a cell 162 for goal amount of die grind, a cell 164 for goal amount of punch grind, and a cell 166 for goal hits per die set. Benchmark data 176 values are included automatically from New Die Data Entry Sheet 10, but the user can enter different benchmark values 176 onto Die Performance Sheet 170. Performance results 178 are determined from, for example, Die Service Record 80 information. Performance results 178 include, for example, cells for total amount removed from die 180, total amount ground from punch 182, average number of hits per 0.001 die grind 184, average number of hits per 0.001 punch grind 186, average number of hits per die set 188, average number of hits per grind 190, average amount of die grind 192, and average amount of punch grind 194. Cells 196 for efficiency ratings are determined by comparing performance results 178 to benchmark data 176. Efficiency ratings include, for example, a ratio 198 of actual average hits per grind to goal hits per grind 160, a ratio 200 of actual average number of hits per 0.001 die grind to benchmark data 176, a ratio 202 of actual average number of hits per 0.001 punch grind to benchmark data 176, a ratio 204 of actual average number of hits per die set to goal hits per die set 166, a ratio 206 of actual average amount removed from die during routine service to benchmark data 176 (i.e. stock removal rate from die), and a ratio 208 of actual average amount removed from punch during routine service to benchmark data 176 (i.e. stock removal rate from punch). Die Performance Sheet 170 summarizes and analyzes a plurality of data from, for example, New Die Data Entry Sheet 10 and Die Service Record 80 and thus provides insights into die performance relative to general industry-accepted benchmarks. For example, ratio 198 of average hits per grind to goal hits per grind 160 provides insight into whether a die is being ground prematurely based on industry norms, whether a die is being serviced correctly and at appropriate intervals, and whether punch or die washout is occurring. Ratio 200 of hits per 0.001 die grind to benchmark data 176 provides insight into die lower half consumption rate and whether die washout is occurring. Ratio 202 of hits per 0.001 punch grind to benchmark data 176 provides insight into die upper half (i.e. punch) consumption rate and whether punch washout is occurring. Ratio 204 of hits per die set to goal hits per die set 166 provides insight into how well a die is designed, built, maintained and suited to press capacity. Hits per die set ratio 204 is affected by, for example, poor die maintenance, punch press overload, improper grinding procedures and off-center punch press loading. Stock removal rate from die ratio 206 and stock removal rate from punch ratio 208 are factors providing insight into punch and die consumption rates and whether sharpening is inadequate or excessive. Stock removal rate from die ratio 206 is a ratio of (total number of hits to date 214 divided by total amount removed from die 180) to (goal hits per grind 160 divided by goal amount of die grind 162). Stock removal rate from punch ratio 208 is a ratio of (total number of hits to date 214 divided by total amount removed from punch 182) to (goal hits per grind 160 divided by goal amount of punch grind 164). FIG. 7 illustrates a Die Summary Sheet 220, which supplies, in summary form, cells for punch life 222 and die life 224 in terms of weeks remaining. Die Summary Sheet 220 includes, for example, a list 226 of dies at plant location 14 and is automatically updated when a Die Service Record 80 is updated. Die Summary Sheet 220 also includes cells 228, 230, 232 and 234 for checks for partial and full punch and die rebuild requirements. Cells 228, 230, 232 and 234 automatically determine need for and display a warning 236 for such rebuild requirement. The user can activate macro button 238 for a die identification number 12 in list 226 to view Die Service Record 80 corresponding to die identification number 12. The above-described method provides a range of useful information, from aspects of a particular servicing occasion to long-term tool performance and efficiency in a plurality of locations. The user thereby gains a broadened perspective on tool consumption and servicing that can serve to inform the user""s selection of tools and suppliers. User access to tool information is enhanced by installing the die evaluation program for use over a computer network, for example, an intranet or internet. The tool data in spreadsheet format also is amenable to further analysis with available spreadsheet methods, for example, to uncover hidden costs associated with poor tool maintenance or operating procedures. The above-described computer-implemented method is modifiable in a plurality of aspects and is applicable for tracking many items in addition to punch/die tooling. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims. |
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claims | 1. A target grid assembly for employment in a target assembly used to produce radioisotopes by bombarding a target material contained in the target assembly with a particle beam, the target assembly further including a target body defining a target reservoir for receiving the target material and a target window for sealing the target reservoir, said target grid assembly comprising: a vacuum window; a target grid defining a target grid portion, a helium input and a helium output, said target grid portion defining a plurality of target grid supports which are configured to form a plurality of target grid oblong openings, said target grid portion defining an upstream side and a downstream side, said vacuum window being positioned against said upstream side, the target window being supported between said downstream side and the target body, a helium space being defined by said plurality of target grid oblong openings between said vacuum window and the target window and being configured such that helium is injectable into said helium space via said helium input and extractable from said helium space via said helium output; and an interception grid positioned upstream of said vacuum window, said interception grid defining an interception grid portion which defines a plurality of interception grid supports, said interception grid supports being configured to form a plurality of interception grid oblong openings, said vacuum window being positioned between said interception grid supports and said target grid supports. 2. The target grid assembly of claim 1 further including a slotted O-ring defining a first slot and a second slot, said first slot being spaced apart from said second slot such that said first slot coincides with said helium input and said second slot coincides with said helium output, said first slot extending perpendicular to an upper portion of each of said plurality of target grid oblong openings and said second slot extending perpendicular to a lower portion of each of said plurality of target grid oblong openings such that helium injected into said helium input flows into said upper portion of each of said plurality of target grid oblong openings thereby filling said helium space and flows out of said lower portion of each of said plurality of target grid oblong openings into said second slot and through said helium output thereby establishing a helium cooling regime. claim 1 3. The target grid assembly of claim 1 wherein said interception grid supports and said target grid supports are aligned. claim 1 4. A target grid assembly for employment in a target assembly used to produce radioisotopes by bombarding a target material contained in the target assembly with a particle beam, the target assembly further including a target body defining a target reservoir for receiving the target material and a target window for sealing the target reservoir, said target grid assembly comprising: a vacuum window; and, a target grid defining a target grid portion, a helium input and a helium output, said target grid portion defining a plurality of target grid supports which are configured to form a plurality of target grid oblong openings, said target grid portion defining an upstream side and a downstream side, said vacuum window being positioned against said upstream side and the target window being supported between said downstream side and the target body, a helium space being defined by said plurality of target grid oblong openings between said vacuum window and the target window; a slotted O-ring defining a first slot and a second slot, said first slot being spaced apart from said second slot such that said first slot coincides with said helium input and said second slot coincides with said helium output, said first slot extending perpendicular to an upper portion of each of said plurality of target grid oblong openings and said second slot extending perpendicular to a lower portion of each of said plurality of target grid oblong openings such that helium injected into said helium input flows into said upper portion of each of said plurality of target grid oblong openings thereby filling said helium space and flows out of said lower portion of each of said plurality of target grid oblong openings into said second slot and through said helium output thereby establishing a helium cooling regime; and, an interception grid positioned upstream of said vacuum window, said interception grid defining an interception grid portion which defines a plurality of interception grid supports, said interception grid supports being configured to form a plurality of interception grid oblong openings, said vacuum window being positioned between said interception grid supports and said target grid supports, said interception grid supports and said target grid supports being aligned. |
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055815863 | abstract | In a drive device of control rod drive mechanisms, a large number of control rod drive mechanisms are divided into a plurality of groups, and there are provided, for each group, a control rod changeover device, and an inverter power source and inverter controller constituting the drive power source. The drive device comprises a control unit that receives control rod position signals from each of the control rod drive mechanisms and that outputs control signals to the control rod changeover device and inverter controller, and a man-machine device constituting an interface with the operator, that outputs control rod drive information to this control unit. |
062917366 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT According to the present invention, leachable lead in treated materials is decreased to levels well below 5.0 mg/l, measured by TCLP test criteria. Waste and process materials having TCLP lead level in excess of 5 mg/l are considered hazardous and must be treated to be brought into compliance with regulatory requirements. Other metal-bearing materials having leachable metals may also be treated according to the present invention to achieve acceptable metal levels. The treatment technology according to another embodiment of the present invention consists of a two step process for treating contaminated soils and/or solid waste materials with chemical treating agents that convert leachable lead to synthetic (man-made) substantially insoluble lead mineral crystals. As used here, "substantially insoluble" means the leachable lead content in the treated waste sample is less than 5.0 mg/l in the extract by the TCLP Test. Another preferred embodiment of the present invention consists of applying technical grade phosphoric acid (TGPA) that contains sulfate as an impurity to leachable and soluble radionuclides and other radioactive substances often found in debris, soils and solid materials. The addition of water aids in the dispersion and percolation of TGPA throughout the contaminated host matrix. Water can be added at any point of the process, either before or after the TGPA addition, or by diluting the TGPA and applying the dilute TGPA to the target matrix. Mixing of the TGPA with the host matrix is optional, dependent upon the permeability and porosity of the host material. When employed, mixing enhances the uniformity of reagent dispersion through the host material. Treatment Chemicals and Additives The treatment chemicals useful in the present invention may be divided into two groups. The addition of water with the additives may facilitate the ultimate mixing and reaction. A first group, "group one", comprises a source of sulfate, hydroxide, chloride, fluoride and/or silicates. These sources are gypsum, lime, sodium silicate, cement, calcium fluoride, alum and/or like similar products. The second group, "group two", comprises a source of phosphate anion. This group consists of products like phosphoric acid (phosphoric), pyrophosphates, triple super phosphate, trisodium phosphates, potassium phosphates, ammonium phosphates and/or similar compounds capable of supplying a phosphate anion. The first step of this novel process comprises the reaction of leachable lead in contaminated soils or solid waste materials with a gypsum powder, calcium sulfate dihydrate (CaSO.sub.4.2H.sub.2 O). Calcium sulfate dihydrate powder reacts with leachable and mobile lead species in wastes to form hard sulfates, which are relatively insoluble in water. In the invention, the powder form of dry calcium sulfate dihydrate, or gypsum, is preferred for blending with lead contaminated materials because it provides a uniform cover or dry coating over the surfaces of the waste particles and aggregates under low moisture conditions. The greatest benefit and fastest reaction is achieved under conditions wherein 95% of the powder is passable through a 100 mesh sieve, and the remaining 5% is passable through a 20 mesh sieve. The amount of gypsum powder employed is typically from 0-30 percent of the weight of solid waste material being treated. The actual amount employed will vary with the degree and type of lead contamination in the waste material or soil, and with the initial composition as well as the condition of the waste material, among other factors. Alternatively, sulfuric acid, or alum in liquid or powder form can also be used as sources of sulfate ion in certain solid wastes that contain sufficient calcium prior to treatment. Treatment Method At lease one component from group one is added to the mixing vessel or reactor, preferably as a dry powder, although slurries could be pumped under certain circumstances. At least one component from group two is added to the mixing vessel or reactor as either liquid reagent or as granular solid materials. The ingredients of group one and group two can be added to the hazardous waste materials simultaneously, and they are pre-mixed and added in a single step. Alternatively, the components of group one and two can be added sequentially in a two-step process with either component added first. That is, the two steps may occur in any order. At least one ingredient of group one can be added in step I or step II. Likewise, at least one ingredient of group two can be added in either step I or step II. Enough water may be added to allow good mixing to prevent dust formation, and to permit good chemical reaction. Not too much water is added to solid materials if the treated waste is to pass the paint filter test. In the first step of the instant process, a thorough and uniform mixing of gypsum powder with the solid waste is accomplished by mixing shredded and screened waste particles with small gypsum particles in, for example, a grizzly or other mixing device. The calcium ions from the gypsum powder displace lead from soil complexes and organic micelles present in the contaminated soil and solid waste material The following equations (1) and (2) describe the reaction of leachable lead with gypsum. ##STR1## The reaction of lead with gypsum forms a "hard sulfate" which crystallizes into mineral species of the barite family--anglesites and calcium-substituted anglesites--which are insoluble in water. The solubility product of lead sulfate is 1.8.times.10.sup.-8, indicating that anglesite crystals would continue to develop over the geologic periods. In the second step of the process, the solid waste material as amended with gypsum powder is treated with a phosphate-supplying reagent, such as (for example), phosphoric acid. Upon contact with the soil or solid waste, the phosphate-supplying reagent reacts chemically to immmobilize the remaining leachable lead. The phosphate-supplying reagent includes phosphate ion sources having one or more reactive phosphate ions, such as phosphoric acid, trisodium phosphate, a potassium phosphate and monobasic or dibasic calcium phosphates. The quantity of phosphate-supplying reagent employed will vary with the characteristics of the solid waste being treated, including particularly such factors as leachable lead content, total lead, and buffering capacity, among other factors. It has been determined that in most instances a quantity of phosphoric acid up to 30 percent of the weight of the waste material is sufficient. The concentration of phosphoric acid in solution will typically range from about 2 to 75 percent by weight. The solution and treatment process are maintained above 30.degree. F. to permit the handling of the phosphoric acid as a liquid reagent. Below 30.degree. F., the phosphoric acid tends to gel while water freezes to form ice, thus creating material handling problems. Free lead, along with calcium ions found in the solid waste (including those imparted through the first step of the process), reacts with the phosphate to form insoluble superhard rock phosphates or calcium substituted hydroxy lead Apatites as shown in Equations (3a) and (3b): ##STR2## The phosphate ions are added to the contaminated soils in solution form; for example, phosphoric acid may be added to water in amounts ranging from about 2 percent to about 75 percent by weight. Phosphoric acid decomposes carbonates and bicarbonates in wastes leading to the synthesis of Apatites and evolution of carbon dioxide gas. Destruction of carbonates and bicarbonates contributes to desirable volume reductions. Although water molecules are generated during the carbonate and bicarbonate decomposition process, it is preferred to have soil moisture at about 10 per cent to about 40 per cent by weight of the soil in order to accelerate the fixation of the leachable lead with the phosphate ions. At this moisture range, material handling is also easy and efficient. It is apparent from Equations (2), (3a) and (3b) that gypsum and phosphoric acid decompose carbonates and bicarbonates during synthesis of new stable minerals of the barite, apatite, and pyromorphite families in soils (as shown in Table I). Decomposition of carbonates and bicarbonates is usually associated with the evolution of carbon dioxide, formation of hydroxyl group, (OH--), and the release of water molecules. As the water evaporates and carbon dioxide molecules are lost to the atmosphere, the treated waste mass and volume are decreased significantly. The solid sulfate powder and the phosphate-supplying reagent are added to contaminated soil and solid waste material having a typical moisture content ranging from about 10 percent to about 40 percent by weight. At a moisture level within the foregoing range, the curing time of the treated materials is approximately 4 hours, which provides adequate time for chemical reactions to occur and immobilize the leachable lead species. Crystals of various lead mineral species begin to form immediately, but will continue over long time periods with an excess of treatment chemicals present. This contributes to "self-healing," as noted during treatability studies as well as full scale treatment operations. Under the foregoing conditions, the immobilization of leachable lead occurs in a relatively dry environment because no wet byproducts, slurries or wastewater are produced by the process of the present invention. Operation of the process under relatively dry conditions beneficially allows cost-efficient handling of the contaminated soils and the waste materials. This allows compliance with Paint Filter Test for solid wastes required by USEPA and RCRA approved solid waste landfill facilities. Effective mechanical mixing, as by a pug mill or other such mixing device, eliminates the need for diffusion in the nonaqueous solid waste matrix. The water resistant and insoluble lead minerals synthesized in soils and solid wastes according to this process are stable, and would behave like naturally occurring rock phosphates and hard sulfates. A list of these synthetic lead mineral species and complexes is presented in Table I below, in order of the relative abundance found during characterization of treated soil by x-ray florescence spectrometry, polarized light microscopy (PLM) and scanning electron microscopy (SEM). TABLE I SYNTHETIC MINERAL SPECIES OF LEAD DETECTED IN A TREATED SAMPLE (LISTED IN DECREASING ORDER OF ABUNDANCE) 31-41% Calcium Substituted Hydroxy Lead Apatites, Ca.sub.0.5-1.5 Pb.sub.3.5-4.5 (OH)(PO.sub.4).sub.3 28-29% Mixed Calcium Lead PHosphate Sulfates, Ca.sub.0.05-0.2 Pb.sub.0.8-0.95 (PO.sub.4).sub.0.15-0.5 (SO.sub.4).sub.0.25-0.75 20-22% Mixed Calcium Anglesites, Ca.sub.0.05-0.3 Pb.sub.0.7-0.95 (SO.sub.4) 3-6% Anglesites, PbSO.sub.4 2-7% Lead Hydroxy/Chlor Apatite, Pb.sub.5 (PO.sub.4).sub.3 (OH).sub.0.5 Cl.sub.0.5 1-3% Pyromorphite, Pb.sub.3 (PO.sub.4).sub.2 1-2% Organo-Lead Phosphate Sulfate, Humus-o-Pb.sub.3 (PO.sub.4)(SO.sub.4) Some of the chemical reactions that occur during the curing stage, and lead to the development of mixed minerals containing both sulfates and phosphates, are illustrated in Equations (4) and (5). ##STR3## The process of the present invention beneficially decreases the volume of the waste materials through: (i) the evolution of carbon dioxide during the chemical decomposition of carbonates and bicarbonates, upon reaction with the acidic components in gypsum and phosphoric acid, and (ii) hardening and chemical compaction as a result of the synthesis of new minerals which result in changes in interstitial spaces and interlattice structures. Applications of the process on a lead contaminated soil was associated with pore space decrease from 38.8% to 34.3% by volume. A decrease in pore space was associated with increased compaction of the treated soils and a decrease in overall waste volume ranging from 21.4% to 23.0%. For different waste types, the volume decrease varies with the amount of treatment chemicals used in the process. In another lead toxic solid waste, application of this process resulted in a volume decrease of the order of 36.4% while decreasing the leachable lead to levels below the regulatory threshold. This reduction in volume of the contaminated soil and the solid waste material makes the process of the present invention particularly beneficial for off-site disposal in a secured landfill by cutting down the costs of transportation and storage space. The process can be accomplished at a cost-efficient engineering scale on-site or off-site for ex-situ treatment of lead-toxic solid wastes. This innovative treatment technology also offers a great potential for in-situ application to immobilize lead most economically without generation of any wastewater or byproducts. FIG. 3 illustrates schematically the process of the present invention. The lead-contaminated uncontrolled hazardous waste site 10 with lead-toxic wastes is subject to excavation and segregation 20 of waste piles based on their total lead and TCLP lead contents into (a) heavily contaminated pile 30A, (b) moderately contaminated waste pile 30B and (c) least contaminated waste pile 30C. The staged soil and solid waste material in piles 30A, 30B and 30C is subjected to grinding, shredding, and screening 50 through an appropriately sized mesh sieve. The screening yields particles that are usually less than 5 inches in diameter for mixing with gypsum powder 40 in a grizzly that allows a uniform coating of gypsum over the soil particles and waste aggregates during the grinding, shredding and/or mixing step. Alternatively, as shown by the dashed line, gypsum powder 40 may be added continuously to the screened solid waste material in prescribed amounts as determined during treatability trials. Most of the leachable lead binds chemically with gypsum at molecular level to form lead sulfate, which crystallizes into a synthetic nucleus of mixed calcium anglesite and pure anglesite minerals identified in the treated material by chemical microscopy techniques. The gypsum-coated waste particles and aggregates are then transported on a belt conveyor 70 or other conveying means to an area where an effective amount of phosphoric acid solution 80 of specified strengths in water 90 is added or sprayed just prior to thorough mixing in a pug mill 100 (or other mixing means). The temperature of the phosphoric solution is preferably maintained above 30.degree. F. to prevent it from gelling. The treated soil and wastes are subject to curing 110 and drying 120 on a curing/drying pad, or may less preferably be cured and dried using thermal or mechanical techniques. The end product of the process passes the Paint Filter Test. During the curing time of about four hours, various "super-hard phosphate" mineral species, such as calcium-substituted hydroxy lead-Apatites and mixed calcium-lead phosphate-sulfate mineral species, are formed in treated waste media 130. Crystals of these mineral species (in early stages of development) have been identified in the treated soil materials and solid wastes by geo-chemical and microscopy techniques like PLM and SEM. The proportions of waste materials and reagents used in the process may be varied within relatively wide limits. For example, the amount of gypsum powder should be sufficient to produce lead sulfate in contaminated solid or solid waste material. In addition, the amount of phosphate-supplying reagent is prescribed in an amount sufficient to produce mineral species such as hydroxy-lead apatite in contaminated soil or solid waste material during a relatively short curing time of 4 hours, usually ranging from about 3 to about 5 hours. Further drying of the treated material may take 24 to 96 hours, but has not been required in any application to date. Table II documents the optimum curing time of 4 hours for the process. In all instances, the leachable lead as measured by the EP Toxicity Test Procedure was found below 0.6 mg/l and the differences between analytical values below this level one statistically insignificant. TABLE II DOCUMENTATION OF OPTIMUM CURING TIME USING EP TOXICITY TEST CRITERIA FOR LEAD FIXATION EP Toxic Pb Concentration in mg/l EP Toxic Pb found Waste in processed sample (Untreated at a Curing Time of Matrix Sample) 4 Hrs. 48 Hrs. 96 Hrs. Catergory mg/l mg/l mg/l mg/l Pb Toxic Soil A 495 0.4 0.4 0.6 Pb Toxic Soil B 46 0.3 0.2 0.2 Pb Toxic Soil C 520 0.3 0.5 0.5 The amount of the gypsum powder and the phosphoric acid employed will be dependent on the amount of contaminant present in the soil, initial characteristics of the solid waste material, whether the material is in-situ or is excavated and brought to an off-site facility for treatment; the same is true for other sulfate compounds and phosphate reagents. The following Example I describes various ratios of the chemical reagents for application to the excavated lead-contaminated solid wastes in order to render the leachable lead substantially insoluble; i.e., to reduce the leachable lead to levels below 5.0 mg/l by EP Toxicity Test lead and TCLP Test criteria now in force under current land-ban regulations. When the present invention is used to treat radionuclides and other radioactive materials, the amounts of treatment chemicals added are a function of the contaminated host matrix geochemistry, the concentration of radionuclides in the host matrix, and the presence of potential interferences that could inhibit the reactions, and the geotechnical properties of the host material. A preferred rate of TGPA addition is in the range of 0.1 to 20% by weight of the matrix to be treated. Preferred water content will also vary with the characteristics of the host material to be treated, but should be in the range of 5% to 50% by weight. Water content may affect the rate of reaction with lower water content requiring longer reaction periods and increased need for supplemental mixing. Higher water content, on the other hand, may adversely impact subsequent material handling, and volume reduction results. Water supplied to an excess will yield a material that will contain free liquids. In these cases, the treated material should be allowed to react for a longer period of time to permit a decrease in moisture content by capillary drying and/or evaporation. In some instances, dewatering or other drying techniques may be used to form a material that contains no free liquids. When TGPA is not utilized as the group two treatment chemical reagent, other compounds that provide soluble phosphates, or phosphates that can be solubilized may be substituted. The phosphates may be applied in a liquid form or as a solid. Prior to employing the process of the present invention at a site, laboratory tests should be conducted to determine the amounts of group one and group two treatment chemicals that will be needed for the contaminated matrix that is to be treated. Identification of carbonates, borates, sulfates, silicates and/or phosphates in the host material will facilitate the selection of the optimum quantities of treatment chemicals. Temperature and Pressure Ambient temperature and pressure may be used for the disclosed treatment process, permitted the operations of the feeding and mixing equipment allow such. Under sub-freezing conditions, phosphoric acid may be heated to 50.degree. F. to prevent it from gelling and in order to keep it in a pumpable viscosity range. Treatment System Design The treatment may be performed under a batch or continuous system of using, for example, a weight-feed belt or platform scale for the metal-hazardous waste materials and a proportionate weight-belt feed system for the dry ingredient or ingredients and powders of at least one of the groups. A metering device, e.g., pump or auger feed system, may instead, or additionally, be used to feed the ingredients of at least one of the groups. The same equipment used for treating metal-hazardous waste material is used for treating soils and waste materials contaminated with radionuclides and other radioactive substances. EXAMPLE 1 Single Step Mixing of Treatment Chemicals A lead contaminated soil from a battery cracking, burning, and recycling abandoned site was obtained and treated with group one and group two chemicals in one single step at bench-scale. The contaminated soil contained total lead in the range of 11.44% to 25.6% and TCLP lead in the ranged of 1781.3 mg/l to 3440 mg/l. The bulk density of contaminated soil was nearly 1.7 g/ml at moisture content of 10.3%. The contaminated soil pH was 5.1 with an oxidation reduction potential value of 89.8 mV. To each 100 g lot of lead hazardous waste soil, sufficient amounts of group one and group two treatment chemicals and reagents were added as illustrated in Table III, in order to render it nonhazardous by RCRA (Resource Conservation and Recovery Act) definition. TABLE III TCLP Test Run Treatment Additive(s) Lead(mg/') I 5% lime, 5% gypsum, 10.2% phosphoric 0.5 II 12% phosphoric, 10% potassium sulfate 2.2 III 12% phosphoric, 10% sodium sulfate 3.5 IV 15% TSP 3.7 V 12% phosphoric, 10% Portland Cement I 0.2 VI 12% phosphoric, 10% Portland Cement II 0.9 VII 12% phosphoric, 10% Portland Cement III 0.3 VIII 12% phosphoric, 10% gypsum 4.6 IX 15% TSP, 10% Portland Cement 0.1 X 15% TSP, 10% Portland Cement II 0.2 XI 15% TSP, 10% Portland Cement III 0.2 XII 15.1% phosphoric 3.6 XIII 10% trisodium phosphate, 10% TSP 1.2 XIV 6.8% phosphoric, 4% TSP 4.5 XV 10% gypsum 340 XVI 12% phosphoric, 5% lime 0.9 Control Untreated Check 3,236.0 It is obvious from TCLP lead analyses of fifteen test runs that the single step mixing of at least one component of either or both group one and group two treatment chemicals is very effective in diminishing the TCLP lead values. In test run I, mixing of lime and gypsum from group one additives and phosphoric from group two decreased the TCLP lead to levels below 1 mg/l from 3440 mg/l with a curing time of less than 5 hours. Although the treatment chemicals of group two are more effective in decreasing the TCLP lead than the treatment chemicals of group one, as illustrated by the comparison of test runs XII and XV for this waste soil, but the combined effect of both groups is even more pronounced in decreasing the leachable lead. Results of these bench-scale studies were confirmed during engineering-scale tests. Single step mixing of 5% lime, 11.76% phosphoric acid and 15% water in a 2000 g hazardous soil diminished the TCLP lead values form 3440 mg/l to 0.77 mg/l in less than 5 hours. Likewise, single step mixing of 300 g Triple Super Phosphate (TSP), 200 g Portland Cement (PC) and 300 ml water in 200 g hazardous soil decreased the TCLP lead to levels below 0.3 mg/l within a relatively short curing time. Single step mixing of both groups of treatment chemicals can dramatically reduce treatment costs making this invention highly attractive and efficient for commercial use. The first advantage of using lime and phosphoric acid combination over the use of TSP and PC is that in the former a volume decrease of 6% was realized when compared to the original volume of untreated material. In the later case, a volume increase of 37% was measured due to hydration of cement. The second advantage of using phosphoric and lime combination is that the mass increase is less than the mass increase when TSP and PC are added. Quantitatively, the mass increase in this hazardous waste soil treatment was approximately 16.7% due to combination of lime and phosphoric whereas the mass increase was about 40% due addition of TSP and PC. And therefore, those skilled scientists and engineers learning this art from this patent, must make an economic judgment for each lead contaminated process material and waste stream which chemical quantity from each group would be most effective in rendering the treated material non-hazardous. The third advantage in using lime and phosphoric over the use of TSP and PC is that the former does not change in physical and mechanical properties of original material and if a batch fails for shortage of treatment chemicals, it can be retreated rather easily by adding more of the treatment reagent. The material treated with PC hardens and may form a monolith which is difficult to retreat in case of a batch failure. EXAMPLE 2 Interchangeability of Two Step Mixing Method In the lead contaminated soil from the abandoned battery recycling operations, the treatment chemicals of either group can be added first and mixed thoroughly in an amount sufficient to decrease the TCLP lead below the regulatory threshold. Two step mixing method of the group one and group two treatment additives is as effective as single step mixing of same quantity of treatment chemicals selected from group one and group two. Table IV illustrates data that confirm that the application of group one treatment chemicals in step I is about as effective as application in step II. The same is true for group two treatment chemicals. Thus, the two steps are essentially interchangeable. The reversibility of the steps according to the present invention make it very flexible for optimization during commercial use, scaling up and retreatment of any batches that fail to pass the regulatory threshold criteria. TABLE IV TREATMENT ADDITIVES TWO STEP MIXING METHODS TCLP TEST TOTAL LEAD RUN STEP I STEP II LEAD mg/l I 10% gypsum & 2% 12% phosphoric acid 20.8 1.8 lime (Group I) (Group II) II 12% phosphoric 10% gypsum & 2% 24.4 1.9 (Group II) lime (Group I) III 10% gypsum 10.6% phosphoric 24.4 3.4 (Group I) (Group II) IV 10.6% phosphoric 10% gypsum 22.4 3.5 (Group II) (Group I) Single Step Mixing Method V 10% gypsum and 12% phosphoric 23.6 3.5 Untreat- Control/Check 23.1 3440 ed EXAMPLE 3 Retreatability and Single Step Mixing A sample of hazardous cracked battery casings of 1/2"-1" size containing 14% to 25.2% total lead and about 3298 mg/l of TCLP was obtained for several test runs of the invention to verify the retreatability of batches that fail because of the insufficient dose of treatment chemical added. The results of initial treatment and retreatment are presented in Table V and compared with single step mixing treatment additives from both groups. About 200 g of hazardous material was treated with 10.5% phosphoric acid, 2.5% gypsum and 1.25% lime, all mixed in one single step. The TCLP lead was decreased from 3298 mg/l to 2.5 mg/l as a result of single step mixing in test run V (TABLE V). When the amount of additive from group two was less than the optimum dose needed, the TCLP lead decreased from 3298 mg/l to: (i) 1717 mg/l when 4.2% phosphoric and 1% lime were added during step I and II respectively, and (ii) 2763 mg/l when 4.2% phosphoric and 5% gypsum were added, compared to untreated control. Since the TCLP lead did not pass the regulatory criteria of 5 mg/l, treated material from test runs I and II were retreated during test runs III and IV, respectively, using sufficient amounts to phosphoric acid (an additive from group two) in sufficient amount to lower the TCLP lead to 2.4 mg/l and 2.5 mg/l, respectively. Furthermore, this example confirms that lime is more effective in decreasing TCLP lead than gypsum among different additives of group one. And as a result, the requirement of group two treatment reagent is lessened by use of lime over gypsum. The example also illustrates that one or more compounds of the same group can be used together to meet the regulatory threshold limit. TABLE V TREATMENT ADDITIVES TWO STEP MIXING METHODS TCLP Lead Test Run Step I Step II mg/l I 4.2% phosphoric 1% lime 1717 II 4.2% phosphoric 5% gypsum 2763 Untreated 3296 Control Retreatment (Single Step Mixing) Method III-I 6.8% phosphoric 2.4 IV-II 8.5% phosphoric 3.5 Single Step Mixing V 10.5% phosphoric, 2.5% gypsum, 2.5 1.25% lime EXAMPLE 4 Wide Range of Applications and Process Flexibility in Curing Time Moisture Content and Treatment Operations TABLE VI illustrates different types of waste matrix that have been successfully treated employing the one step and two step mixing treatment additives from group one and group two. For these diverse waste types and process materials, total lead ranged form 0.3% to 23.5%. This example discloses the flexibility and dynamics of the treatment process of the invention in rendering non-hazardous, by RCRA definition, a wide range of lead-hazardous and other metal-hazardous materials within a relatively short period of time, usually in less than 5 hours. It is expected that this process will also render bismuth, cadmium, zinc, chromium (III), arsenic (III), aluminum,, copper, iron, nickel, selenium, silver and other metals also less leachable in these different types of wastes. The moisture content of the waste matrix is not critical and the invented process works on different process materials and waste types independent of the moisture content. The treatment operations can be carried out at any level--bench, engineering, pilot and full-scale--on relatively small amounts of hazardous waste material in laboratory to large amounts of contaminated process materials, soils, solid wastes, waste waters, sludges, slurries and sediments outdoor on-site. The process is applicable in-situ as well as ex-situ. TABLE VI UNIVERSE OF APPLICATION FOR THE INVENTION MACTITE TREATMENT PROCESS LEAD LEACHABLE LEAD (mg/l) CONTAMINATED TREATMENT TOTAL Before After VOLUME WASTE TYPE ADDITIVE LEAD % Treatment Treatment DECREASE OLD DIRT 3.4% Phosphoric 2.2 164.4 1.5 16.7 WASTE WITH BROKEN 8.1% Lime 2.7 197.5 ND (<.5) BATTERY CASING 1% Gypsum and 3.4% Phosphoric SLAG-LEAD SHELTER 10.2% Phosphoric 6.6 21.3 2.0 LEAD-BIRD SHOT 16% Phosphoric 16.1 3720 ND (<.5) 14% Lime and 30% Gypsum LEAD-BUCK SHOT 16% Phosphoric 11.4 1705 ND (<.5) 14% Lime and 28% Gypsum BATTERY CASINGS 5% Gypsum 12 288 0.6 0 ORGANIC HUMUS SOIL 0.5% Lime 1.9 23.2 ND (<.5) 29 2.0% Phosphoric 50:50 MIXTURE OF 4% Gypsum 0.5 687 0.7 3.3 CASINGS AND SAND 4% Phosphoric 422.2 0.95 23.6 SOLID WASTE SOIL 3% Lime 23.5 12.0 6.0 Contaminated With 12% Phosphoric Tetraethyl lead SOIL CONTAMINATED 10% Gypsum 4.74 590 13.7 WITH LEADED 6% Phosphoric GASOLINE 3% Lime 3.2 213 1.6 5.1% Phosphoric CARBON WITH 4.7% Phosphoric 12.6 105.6 0.5 LEAD DROSS WIRE FLUFF 1.7% Phosphoric 0.3 19 0.7 WIRE CHIP 0.75% Phosphoric 0.4 28 ND (<.2) LAGOON SEDIMENT 0.6% TSP 0.3 3.9 0.23 0.5% Phosphoric 5.6 0.3 RCRA ORGANIC SLUDGE 0.6% Phosphoric 9.4 580 ND (<.5) 10% Gypsum FILTER CAKE 8.5% Phosphoric 2.9 245.3 1.1 GRAVEL 5% Gypsum 0.16 7.5 0.5 2.2% Phosphoric ROAD GRAVEL 10% Gypsum 0.34 46 ND (<.5) 8.4% Phosphoric MIXTURE OF BATTERY 2.5% Gypsum 1.3 75 0.6 19.6 CASINGS (SOLD WASTE) 3.4% Phosphoric AND SOIL INDUSTRIAL WASTE 1 g Lime 2.75 91 0.7 (B) 3.4% Phosphoric INDUSTRIAL PROCESS 3.4% Phosphoric 1.3 61 ND (<.5) MAT. (G) SOIL (B) 3.4% Phosphoric 4.1 129.5 0.6 25.6 SOIL (S) 50% Gypsum 11 <0.01 SOIL (O) 1.3% Phosphoric 0.38 34.6 ND (<.5) SOIL (C) 5% Lime 11.78 130.6 0.33 8.5% Phosphoric BATTERY CASINGS 5% Gypsum 2.5 110.1 1.9 3.4% Phosphoric GRAY CLAY SOIL 5% Trisodium 2.2 46.6 0.2 Phosphate EXAMPLE 5 Nearly twenty (20) different chemicals and products from various vendors and supply houses were screened for chemical fixation of leachable lead in hazardous solid waste samples. Only six (6) of these treatments chemicals were found effective in decreasing the leachable lead as measured by: (1) the EP Toxicity Test and (2) the TCLP Test. Table VII presents a summary of leachable lead found in untreated and treated waste samples allowed to cure for a minimum of 4 hours after treatment with at least one of the effective chemicals. Treatment chemicals found relatively ineffective for lead fixation included a variety of proprietary products from American Colloid Company and Oil Dri, different sesquioxides like alumina and silica, calcium silicate, sodium silicate, Portland cement, lime, and alum from different vendors. Results for these ineffective chemicals are not shown in Table VII. TABLE VII RELATIVE EFFECTIVENESS OF VARIOUS TREATMENT CHEMICALS SCREENED TO DECHARACTERIZE THE LEAD-TOXIC SOLID WASTES Leachable Lead in mg/l Treatment Chemical (Step) Toxicity Test EP TCLP Test I. Untreated Control 221.4 704.5 II. Single Treatment Chemical (One Step Treatment) a. Sulfuric Acid (I) 11.7 39.8 b. Phosphoric Acid (I) 1.0 5.9 c. Superphosphate Granular (I) 2.7 11.4 d. Liquid Phosphate Fertilizer (I) 19.4 64.3 e. Gypsum Powder (I) 24.9 81.8 f. Sodium Phosphate (I) 28.7 93.9 III. Two Step Treatment g. Sulfuric (I) & Lime (II) 20.6 68.1 h. Gypsum Powder (I) & Alum (II) 3.9 15.3 i. Sodium Phosphate (I) & 3.1 12.6 Phosphoric (II) j. Gypsum (I) & Phosphoric (II) N.D.* 1.6 IV. Three Step Treatment k. Gypsum (I), Alum (II) & 12.8 43.3 Sodium Phosphate (III) l. Gypsum (I), Phosphoric (II) & N.D.* 1.4 Sodium Phosphate (III) *N.D. means non-detectable at <0.50 mg/l. Evaluation of a single treatment chemical in one step reveals that phosphoric acid was most effective in fixation of leachable lead followed by granular super-phosphate, a fertilizer grade product available in nurseries and farm supply houses. However, neither treatment effectively treated leachable lead to the USEPA treatment standard of 5.0 mg/l by TCLP methodology. Although both phosphoric acid and granular superphosphate were effective in meeting the now obsolete EP Toxicity Test criteria at 5.0 mg/l, this test has been replaced by TCLP Test criteria for lead of 5.0 mg/l. Single application of the phosphoric acid, granular superphosphate or any other chemical was short of meeting the regulatory threshold of 5.0 mg/l by TCLP Test criteria for lead. In a two-step treatment process, application of gypsum during Step I and treatment with phosphoric acid in Step II resulted in decrease of TCLP-lead consistently and repeatedly below the regulatory threshold of 5.0 mg/l. The results of this two-step treatment process utilizing gypsum in Step I and phosphoric acid in Step II are most reliable and hence, the two-step process may be applied to a wide variety of lead contaminated wastes as exhibited in Example II. A three-step process, as set forth in Table VII, was not perceived to be as economically viable as a two-step treatment process, despite its ability to reduce lead levels in satisfaction of the TCLP Test criteria. A process that employs the beneficial combination of treatment first with a sulfate compound and then with a phosphate reagent in accord with the present invention, in combination with one or more additional treatment steps, may nevertheless be within the scope of the invention. In order to illustrate the relative proportions of two chemicals, e.g., gypsum and phosphoric acid, needed for treatment of lead-toxic wastes, three soil samples from a lead contaminated test site were processed using the present invention, in which gypsum powder was used in the first step, and phosphoric acid solution in water at concentrations of about 7, 15 and 22 percent by weight in the second step. The soil was measured for lead content in accordance with the EP Toxicity Test before and after treatment. A level of leachable lead below 5 mg/l was considered non-hazardous according to this procedure. During these test runs, the EP Toxicity Test criteria were in force for treated waste material. The results of these tests are set forth in Table VIII: TABLE VIII EFFECTIVENESS IN FIXATION AND STABILIZATION OF LEACHABLE LEAD IN LEAD TOXIC SOILS EP TOXIC LEAD PROCESS STEPS TEST REULTS Soil Sample Gypsum Phosphoric Before After (Lead-toxic Step I Step II Treatment Treatment waste) (g/kg soil) (g/kg soil) mg/l mg/l 1. Low lead 20 10 8 <0.1 contamination 2. Moderate 30 20 61 <0.1 contamination 3. Highlead 40 30 3,659 1.7 contamination The foregoing results demonstrate that the process of the present invention was effective in all three samples, representing 3 different levels of lead contamination. The process is flexible and is usually optimized during bench scale treatability studies for each waste type to immobilize the leachable lead and to decharacterize or transform the lead-toxic waste into non-toxic solid waste acceptable to TSD facilities under current land ban regulations. A net reduction of 36.4% in waste volume through use of the instant process has been observed. Typical volume reductions are set forth in Table IX. TABLE IX CHANGES IN SOLID WASTE VOLUME AS A RESULT OF TREATMENT WITH THE TWO-STEP PROCESS SOLID WASTE VOLUME Final (After Decrease in SOLID WASTE Initial (Before Application of Waste MATERIAL Application of Process and Volume (Treatment Scale) Process) Curing) (%) 1. Low toxic soil 3,85O cu. yd. 2,450 cu. yd. 36.4 (full scale) 2. Lead toxic Solid Waste (Bench Scale) Test Run I 106.1 cu. in. 81.51 cu. m. 23.0 Test Run II 22.0 cu. in. 17.3 cu. in. 21.4 The most profound effect of the process of the present invention is at a structural level, where the break-down of granular aggregates is associated with a loss of fluffiness and a decrease in pore space and increased compaction due to physical, mechanical and chemical forces at different levels. At a molecular level, phosphoric acid breaks down the minerals containing carbonates and bicarbonates, including cerussites, in stoichiometric proportions. Soon after the addition of phosphoric acid to a solid waste containing cerussites, extensive effervescence and frothing becomes evident for several minutes and sometimes for a few hours. The phosphoric acid breaks down the acid sensitive carbonates and bicarbonates leading to the formation of carbon dioxide, water and highly stable and insoluble sulfate and phosphate mineral compounds. Thus, structural changes due to interlattice reorganization as well as interstitial rearrangement in waste during processing are associated with an overall decrease in waste volume. Depending on the extent of carbon dioxide loss from the breakdown of carbonates and bicarbonates present in the lead-toxic solid waste, the process may lead to a slight loss of waste mass as well. Water generated during the chemical reactions is lost by evaporation, which further decreases the mass and volume of the treated solid wastes and soils. The cost of the process of the present invention is moderate to low, depending upon (i) waste characteristics, (ii) treatment system sizing, (iii) site access, (iv) internment of final disposition of treated material and (v) site support requirements. The costs of treatment and disposal are presently on the order of $115 per ton of lead-toxic waste, as compared to off-site conventional treatment and disposal costs of over $250 per ton if no treatment in accord with the invention had been performed. Moreover, recent land ban regulations would prohibit the disposal of all lead-toxic wastes in landfills. The foregoing Example makes clear that the process of the present invention provides an efficient technology that is economically attractive and commercially viable in meeting regulatory criteria for landfills. EXAMPLE 6 The process of the present invention was applied on bench scale to five different lead-toxic waste materials that were characterized for total lead, TCLP-lead, moisture content and pH before and after treatment. A curing time of 5 hours was allowed for completion of the treatment process. The results compiled in Table X exhibit the profound effects of the process in decreasing the TCLP lead in a wide range of lead-toxic soils and solid wastes containing total lead as high as 39,680 mg/kg and TCLP lead as high as 542 mg/l. In each of the five cases, the instant process immobilizes the leachable lead to levels below the regulatory threshold of 5 mg/l set by the TCLP Test criteria for lead currently in force under the land ban regulations of the United States Environmental Protection Agency. TABLE X TYPICAL CHANGES IN SOLID WASTE CHARACTERISTICS DUE TO PROCESS EFFECTS MEASURED VALUES SOLID WASTE Before After Treatment CHARACTERISTICS Treatment & Curing I. Lead-toxic SW-A Total lead, % 1.442 1.314 TCLP-Lead, mg/l 542.0 2.0 Moisture, % 23.0 33.0 pH, S.U. 8.1 4.8 II. Lead-toxic SW-B Total lead, % 0.847 0.838 TLCP-Lead, mg/l 192.0 2.4 Moisture, % 27 36 pH, S.U. 8.0 5.3 III. Lead-toxic SW-C Total lead, % 3.968 3.066 TLCP-Lead, mg/l 257.6 1.0 Moisture, % 10.0 18.1 pH, SU. 7.2 4.5 IV. Lead-toxic SW-D Total lead, % 2.862 2.862 TLCP-Lead, mg/l 245.3 0.38 Moisture, % 71.6 84.1 pH, SU 8.1 6.3 V. Lead-toxic SW-E Total lead, % 0.16 0.12 TLCP-Lead, mg/l 7.5 1.87 Moisture, % 12.3 23.0 pH, S.U. 7.0 5.4 It is obvious from Table X that the instant process operates over a wide range of moisture and pH conditions. It is associated with 8 to 11% rise in moisture content. The end product of the treatment process may contain moisture in a typical range of 18% to 36% on a dry weight basis. The end product passes the Paint Filter Test for solids and there are not other byproducts or side streams generated during the process. The treated solid waste is cured in 4 to 5 hours and may be allowed to dry for 2 to 3 days after treatment for loss of unwanted moisture prior to final internment and disposition. This time is sufficient for the TCLP Tests to be completed as part of the disposal analysis under land ban regulations enforced by the USEPA. It is necessary to establish the quantities of gypsum and phosphate reagent on a case-by-case basis, because the consumption of these materials will depend not only upon the initial lead level in the waste or soil, but also upon other waste characteristics such as cation exchange capacity, total buffering capacity, and the amounts of carbonates and bicarbonates present, among others. Bench scale treatability studies for each solid waste considered will be necessary to determine the optimum levels of the materials that are employed. The treatability studies are designed to optimize the amount and grade of gypsum powder (or other sulfate compound) needed during step I, and the amount and concentration of phosphoric acid (or other phosphate compound) needed in step II for cost-effective operation of the treatment system. Those skilled in the art are knowledgeable of such bench studies, which are usually carried out as precursors to full scale treatment. Several series of studies were performed on host matrices containing leachable and soluble radionuclides and other radioactive substances using the present invention. EXAMPLE 7 Sample material from a site in the eastern United States was homogeneously mixed in a container. The material consisted of silts, clays, sand and gravel mixed with glass, nails, rocks and debris. The material was collected from an environmental restoration project where site efforts focused on excavation, packaging, transportation and disposal of Thorium contaminated soil and materials from beneath residential homes. Three 300 g sub-samples of untreated material were prepared from the sample material with the materials in each of the sub- samples sized to less than 3/8 inch and suitable for USEPA SW-846 Method 1311 (TCLP) extraction. Sample 1 (US-1) was extracted using TCLP fluid No. 1, Sample 2 (US-2) was extracted using TCLP fluid No. 2, and Sample 3 (US-3) was extracted using laboratory grade deionized ("DI") water as the only modification to the EPA method. This soil characterization step was conducted for purposes of determining the most harsh extraction conditions for the untreated material. TCLP fluid No. 1 was prepared with glacial acetic acid and 1N NaOH with an end pH of 4.93+/-0.05 S.U. TCLP fluid No. 2 was prepared with glacial acetic acid and deionized water with an end pH of 2.88 +/-0.05 S.U. The laboratory grade DI water had a pH of 6.82+/-0.05 S.U. After tumbling 100 g of the 300 g sub-sample in 200 ml of extraction fluid for eighteen (18) hours at 30+/-2 rpm in a longitudinal rotary TCLP agitator, the extracts were decanted from the settled solids, filtered as per the method, and then placed in Marinelli containers. Radionuclide leachability was determined by conducting total gamma spectroscopy analysis on each extract in accordance with accepted quantification methods using a Nuclear Data Genie Model ND9900 Gamma Spectrometer integrated with a DEC Micro VAX II computer. Each extract was counted for sixteen (16) hours. All results presented below are in the units of picocuries per liter (pCi/l). TABLE XI EASTERN UNITED STATES UNTREATED SAMPLE MATERIAL CHARACTERIZATION US-1 US-2 US-3 .sup. Untreated .sup. Untreated Untreated Radionuclide TCLP Fluid 1 TCLP Fluid 2 Deionized Water Pb-210 329 .+-. 30 173 .+-. 45 175 .+-. 37 Bi-211 2,751 .+-. 736 3,360 .+-. 797 3,451 .+-. 560 Bi-214 772 .+-. 93 1,002 .+-. 120 1,017 .+-. 106 Pb-214 810 .+-. 350 910 .+-. 242 966 .+-. 202 Fr-223 2,183 .+-. 660 3,768 .+-. 73 3,228 .+-. 531 Ra-223 939 .+-. 404 1,514 .+-. 383 714 .+-. 148 Ra-224 1,551 .+-. 503 1,772 .+-. 358 1,868 .+-. 321 Ra-226 1,090 .+-. 167 1,294 .+-. 162 1,352 .+-. 156 Ac-227 213 .+-. 20 243 .+-. 54 173 .+-. 31 Th-227 533 .+-. 163 921 .+-. 179 788 .+-. 131 Th-228 8,335 .+-. 1014 16,490 .+-. 12 13,170 .+-. 1,371 Pa-231 1,136 .+-. 476 1,764 .+-. 467 1,490 .+-. 307 Th-234 22 .+-. 6 19 .+-. 13 10 .+-. 9 U-235 190 .+-. 22 313 .+-. 38 281 .+-. 29 As shown by the gamma spectral analysis of each extract, TCLP fluid No. 2 was identified as the most rigorous extraction fluid for the soil material, primarily because of leachable Thorium and Uranium levels. This fluid was then selected to be used for extraction of the treated samples for the remainder of the studies. In the second portion of the study, two (2) 300 g samples were prepared from the eastern U.S. sample material and labeled as TS-1 and TS-2. Each sample was placed in a laboratory beaker and 35 ml of deionized water and 5% (TS-1) and 10% (TS-2) by weight TGPA were added. The contents in each of the beakers were then mixed by folding with a laboratory spatula in order to simulate blending achievable using full-scale methods in the field. The samples were then allowed to react overnight. Each beaker was then sub-sampled, material particles sized to less than 3/8 inch, and prepared for USEPA SW-846 Method 1311 (TCLP) extraction using 100 g of treated sub-sample material and 2000 ml TCLP Fluid No. 2. Table XII presents the data from the gamma spectral analysis with all units reported as pCi/l. The results from Table XI for untreated materials extracted using TCLP Fluid No. 2 were used as a control and are shown in the fourth column. TABLE XII EASTERN UNITED STATES SAMPLE MATERIAL TREATED WITH DJ WATER AND TGPA TCLP EXTRACTION FLUID NO. 2 RESULTS Radio- TS-1 TS-2 US-2 nuclide 5% TGPA 10% TGPA TCLP Fluid No. 2 Pb-210 <MDA* <MDA 173 .+-. 45 Bi-211 <MDA <MDA 3,360 .+-. 797 Bi-214 <MDA <MDA 1,002 .+-. 120 Pb-214 <MDA <MDA 910 .+-. 242 Fr-223 <MDA <MDA 3,768 .+-. 73 Ra-223 <MDA <MDA 1,514 .+-. 383 Ra-224 <MDA <MDA 1,772 .+-. 358 Ra-226 <MDA <MDA 1,294 .+-. 162 Ac-227 <MDA <MDA 243 .+-. 54 Th-227 <MDA <MDA 921 .+-. 179 Th-228 <MDA <MDA 16,490 .+-. 12 Pa-231 <MDA <MDA 1,764 .+-. 467 Th-234 <MDA <MDA 19 .+-. 13 U-235 <MDA <MDA 313 .+-. 38 *<MDA = less than the calculated Minimum Detectable Activity for the counted sample MDA is the smallest amount of activity that can be detected in a sample. Data from TS-1 was corroborated by a second laboratory on duplicate sample extract for QA/QC data validation purposes. As indicated by the data from Tables XI and XII, TGPA substantially reduces the leachability of radionuclides in soil as determined by USEPA SW-846 Method 1311 (TCLP) extraction with fluid No. 2 and gamma-spectral analysis of resultant extract. It should be noted that the soil sample materials were not sized to less than 3/8 inch until after the TGPA and deionized water were mixed and allowed to cure overnight. The leaching of Thorium, its decay-daughters, and other radionuclides from untreated material was effectively reduced by the addition of TGPA to the material. The treated material was moist after curing overnight, but contained no free liquids. After exposure to the air for forty-eight (48) hours, the treated material was dry and crumbly with nonuniform cohesivity. Volume reduction was observed, but not quantified. EXAMPLE 8 In another study, samples of the untreated material used in Example 7 were mixed with TGPA and other compounds. For this study, gypsum, calcium oxide, triple superphosphate (TSP), and TGPA were selected based upon a generally desired pH range of the end product. Four 300 g samples were prepared: TS-3=35 ml DI water+8% gypsum+5% TGPA; TS-4=35 ml DI water+9% calcium oxide+8% TGPA, TS-5=35 ml DI water+3% calcium oxide+5% TGPA; and TS-6=45 ml DI water+10% TSP+1.6% calcium oxide. Treatment samples received variable amounts of water so that after mixing, the consistency of the mixtures was uniform for all of the samples and there were no free liquids. The water assisted in the dispersement of the reagent and calcium oxide hydration; and hence, the disassociation of the phosphate to a soluble form. Additional water was required in TS-6 because of the solid reagent forms and the hydration demand of CaO in the presence of dry TSP. Table XIII presents the data from USEPA SW-846 Method 1311 (TCLP) extracts of TS-3, TS4, TS-5, and TS-6 analyzed by total gamma-spectroscopy in accordance with procedures outlined in Example 7. All samples were analyzed with TCLP fluid No. 2 (acetic acid+water with a pH of 2.88+/-0.05 S.U.). TABLE XIII EASTERN UNITED STATES SAMPLE MATERIAL TREATED WITH OTHER EMBODIMENTS TCLP EXTRACTION FLUID NO 2 RESULTS Radionuclide TS-3 TS-4 TS-5 TS-6 Pb-210 <MDA <MDA <MDA <MDA Bi-211 <MDA 180 .+-. 69 296 .+-. 106 <MDA Bi-214 <MDA 55 .+-. 23 75 .+-. 29 <MDA Pb-214 <MDA <MDA 50 .+-. 50 <MDA Fr-223 <MDA <MDA <MDA <MDA Ra-223 <MDA 245 .+-. 97 84 .+-. 34 <MDA Ra-224 <MDA <MDA <MDA <MDA Ra-226 <MDA <MDA 122 .+-. 114 <MDA Ac-227 <MDA <MDA 286 .+-. 47 <MDA Th-227 <MDA <MDA 552 .+-. 31 <MDA Th-228 <MDA <MDA <MDA <MDA Pa-231 <MDA <MDA <MDA <MDA Th-234 <MDA <MDA 139 .+-. 53 <MDA U-235 <MDA <MDA 79 .+-. 35 <MDA *<MDA = less than the calculated Minimum Detectable Activity (MDA) for the counted sample Data from samples TS-3 and TS-6 was corroborated by a second laboratory on duplicate sample extracts for QA/QC data validation purposes. As evidenced by the data, the treatment regimes utilizing gypsum+TGPA, calcium oxide+TGPA, and triple superphosphate (TSP)+calcium oxide resulted in the reduction of nuclide leachability. Each of the treatment regimes provided soluble phosphates, or phosphates that were solubilized by pH manipulation in the presence of a fluid. Each of the treatments resulted in the formation of Apatites within the host material, with mineral crystal nucleation chemically incorporating the radionuclides. EXAMPLE 9 The tests in Example 9 were performed to study the volume change of materials treated by the process of the present invention. In Example 9, soil volume was examined prior to and after the addition of TGPA. Because of the difficulty in examining volume changes due to varied conditions, geometric configuration, and chemical properties of material differing between pre- and post-treatment, a special device was constructed to account for changes in density, moisture content, and geotechnical properties. The test apparatus used for measuring the volume consisted of a removable stainless steel cylindrical cup with a flat bottom ("the cup"). The cup had a 10.3 cm inside diameter and a 29.6 cm inside height and mounted vertically to the base of the test apparatus. Mounted above the cup on the apparatus frame was a pneumatic piston with a 1.4 cm thick plate fixed to the piston shaft. When activated with compressed air, a 10.2 cm diameter close-tolerance plate fixed to the piston shaft extended downward and into the open end of the cup. Compressed air operated the piston and was adjusted with a valve so that from 1 to 100 psi could be exerted on soil placed within the cup. The untreated material from Example 7 was used to prepare ten aliquots (of approximately 100 g) which were individually weighed using a top-loading electronic balance (+/-0.01 g). The ten aliquots were then sequentially emptied into the cup. After the addition of each 100 g aliquot, the cylindrical cup was placed in the apparatus and the piston activated to exert a pressure of 10 psi on the soil column. This procedure was repeated until all ten 100 g aliquots had been added and compacted. The height of the soil column was then determined by measuring from the top of the cup to the top of the plate, correcting for the plate thickness, and subtracting the total from the inside height of the cup. The untreated material was then removed from the cup and placed in a laboratory beaker. Care was taken to ensure all visible material was removed and transferred. Water was added to the beaker on a weight basis equal to 12% of the untreated material. TGPA was then added at a dose of 5%, also by weight, of the untreated material. The untreated material and amendments were mixed with a laboratory spatula by folding and allowed to sit overnight. The treated material was then removed from the beaker and placed in the cylindrical cup in ten stages of approximately 100 g each. The pneumatic piston was activated at the same 10 psi pressure each time treated material was added to the cup. After all of the treated material was transferred and compacted with the apparatus, the resultant column height was calculated as previously described. After the material had been allowed to sit for approximately seven (7) days, the volume test was performed again in the same manner. The results of the study are presented in Table XIV. TABLE XIV VOLUME CHANGE OF EASTERN UNITED STATES SAMPLE MATERIAL TREATED WITH 5% (WT.) TGPA Mass Height Mass Height Mass Height Treated Treated Treated Treated Untreated Untreated <24 hours <24 hours 7 days 7 days grams (cm) (grams) (cm) (grams) (cm) 1003.09 8.2 1074.77 7.4 942.51 6.7 These test results show a total volume reduction of 9.75% after 24 hours and 22.4% after 7 days, relative to the initial untreated material. In the next series of studies, sample material from a site in the Midwestern United States was utilized in treatability studies. The material contained small soil grains (with 100% passing through a 9.5 mm sieve) and was comprised of 30% sand, 47% silt, and 23% clay as determined by ASTM D-422 (Particle-Size Distribution). The average density of the material was 1.43 g/cc and the material had a moisture content of 16 percent by weight and a pH of 6.0 S.U. As in the previous examples, the sample material was characterized for radionuclides and other radioactive substances. Nuclide leachability was examined utilizing the Toxic Characteristic Leaching Procedure (TCLP) extraction procedure (USEPA SW-846, Method 1311). Material was also subjected to other leaching tests including the Synthetic Precipitation Leaching Procedure (SPLP) extraction procedure (USEPA SW 846, Method 1312), and a modified version of the TCLP extraction method, where deionized water was substituted for the extraction fluid (DI/TCLP). Results of the gamma-spectral, Uranium, and Technetium-99 characterization analyses on extraction fluids are presented in Table XV. TABLE XV UNTREATED MIDWESTERN UNITED STATES SAMPLE MATERIAL RADIONUCLIDE LEACHABILITY CHARACTERISTICS USA US-5 US-6 Radionuclide/ Method 1311 Method 1312 Modified-1311 Isotope/Item TCLP SPLP DI/TCLP Ra-226 3,644 .+-. 895 3,120 .+-. 494 556 .+-. 219 U-235 266 .+-. 66 190 .+-. 43 39 .+-. 25 U238* 12,308 .+-. 969 11,210 .+-. 92 2,590 .+-. 45 Pb-212 16 .+-. 4 <MDA <MDA Th-234 485 .+-. 138 355 .+-. 90 228 .+-. 73 Tc-99 238 .+-. 11 152 .+-. 10 235 .+-. 11 U 8,698 .+-. 68 7,922 .+-. 65 1,830 .+-. 32 U, total(ug/l) 17,979 16,375 3,783 NOTE: All units in pCi/l, unless indicated *U-238 concentrations were calculated. <MDA = less than the calculated Minimum Detectable Activity (MDA) for the counted sample EXAMPLE 10 In this example, four 400 g samples of soil material (TS-7, TS-8, TS-9 and TS-10) were prepared from the untreated Midwestern U.S. sample material and placed in separate laboratory beakers. Sample TS-7 was used as a control and mixed only with 120 ml of deionized water. For each of the three other samples, 120 ml of deionized water and varying amounts of TGPA were added to each beaker and mixed until a uniform consistency was achieved: TS-8=120 ml DI water+3% (wt.) TGPA; TS-9=120 ml DI water+5% (wt.) TGPA; and TS-10=120 ml DI water+10% (wt.) TGPA. When the mixing was completed, no free liquids were present. After sitting overnight, a 100 g sample of treated material was removed from each beaker and extracted by USEPA SW-846, Method 1311 (TCLP), using Fluid No. 2, to simulate exposure to acidic landfill leachate. The radionuclide leachability for each extract was then quantified by gamma spectroscopy. Total Uranium and Technetium-99 tests were also conducted. Uranium-238 was calculated, assuming the total Uranium present was 100% depleted. The levels of leachable radionuclides and other radioactive substances in the sample material after treatment are presented below in Table XVI. The results in Table XVI can be compared to the results for sample US-4 in Table XV for reference. TABLE XVI RADIONUCLIDE LEACHABILITY OF MIDWESTERN UNITED STATES SAMPLE MATERIAL IN USEPA SW-846, METHOD 1311 (TCLP) FLUID NO. 2 EXTRACT AFTER TREATMENT WITH TGPA Radionuclide/ TS-7 TS-8 TS-9 TS-10 Isotope/Item DI WATER 3% TGPA 5% TGPA 10% TGPA Ra-226 3,114 +/- 568 <MDA <MDA <MDA U-235 231 +/- 55 <MDA <MDA <MDA U238*(ug/l) 5,847 +/- 184 54.5 +/- 1.7 51.7 +/- 1.7 53.5 +/- 1.7 Th-234 230 +/- 97 <MDA <MDA <MDA Tc-99 213 +/- 14.3 67.6 +/- 8.5 55.6 +/- 10.4 3.7 +/- 4.8 U 4,132 +/- 130 38.5 +/- 1.2 36.5 +/- 1.2 37.8 +/- 1.2 U, total(ug/l) 8,541 80 75 78 NOTE: All units in pCi/l, unless indicated *U-238 concentrations were calculated. <MDA less than the calculated Minimum Detectable Activity (MDA) for the counted sample EXAMPLE 11 100 g samples of material treated in Example 10 (TS-7, TS-8, TS-9 and TS-10) were sub-sampled, extracted and analyzed by USEPA SW-846, Method 1312 (SPLP), where the extraction fluid utilized simulated acid rain. Each extract was then quantified for radionuclides by gamma-spectroscopy, and total Uranium and Technetium-99 tests were conducted. Uranium-238 was calculated, assuming the total Uranium present was 100% depleted. The results of the leachable radionuclides and other radioactive substances in the soil after treatment are presented below in Table XVII. The results in Table XVII can be compared to the results for sample US-5 in Table XV for reference. TABLE XVII RADIONUCLIDE LEACHABILITY IN EPA SW-846, METHOD 1312 (SPLP) EXTRACT AFTER TREATMENT WITH TGPA Radionuclide/ TS-7 TS-8 TS-9 TS-10 Isotope/Item CONTROL 3% TGPA 5% TGPA 10% TGPA Ra-226 2,622 +/- 443 233 +/- 136 <MDA <MDA U-235 153 +/- 37 <MDA <MDA <MDA U-238* 6,065 +/- 192 30.1 +/- 1.0 8.8 +/- 0.1 7.3 +/- 0.1 Th-234 170 +/- 81 <MDA <MDA <MDA Tc-99 210 +/- 15 55.6 +/- 7.8 23.2 +/- 6.5 69.8 +/- 7.6 U 4,286 +/- 136 21.3 +/- 0.7 6.3 +/- 0.1 5.2 +/- 0.1 U, total(ug/l) 8,859 44 13.9 10.7 NOTE: All units in pCi/l, unless indicated *U-238 concentrations were calculated. <MDA = less than the calculated Minimum Detectable Activity (MDA) for the counted sample EXAMPLE 12 100 g samples of treated soil material in Example 10 (TS-7, TS-8, TS-9 and TS-10) were subsampled and extracted by USEPA SW-846, Method 1311 with laboratory grade deionized water substituted for the extraction fluid. Although material treated by the invention would never likely be exposed to similar fluid except in the laboratory settings, deionized water is considered by many to be a harsh extraction test as leachable ionic species will tend to diffuse from zones of high concentration to zones of low concentration. Each DI water extract was then quantified for radionuclides by gamma-spectroscopy, and total Uranium and Technetium-99 tests were conducted. Uranium-238 was calculated, assuming the total Uranium present was 100% depleted. The results showing the level of leachable radionuclides and other radioactive substances in the soil after treatment are presented below in Table XVIII for TS-7, TS-8, TS-9 and TS-10. The results in Table XVIII can be compared to the results for sample US-6 in Table XV for reference. TABLE XVIII RADIONUCLDE LEACHABILITY IN EPA SW-846, MODIFIED METHOD 1311 WITH DI EXTRACTION WATER AFTER TREATMENT WITH TGPA Radionuclide/ TS-7 TS-8 TS-9 TS-10 Isotope/Item CONTROL 3% TGPA 5% TGPA 10% TGPA Ra-226 940 +/- 278 <MDA <MDA <MDA U-235 55 +/- 40 <MDA <MDA <MDA U-238* 1,807 +/- 57 30.1 +/- 1.0 8.8 +/- 0.1 7.3 +/- 0.1 Th-234 103 +/- 89 <MDA <MDA <MDA Tc-99 207 +/- 15 55.6 +/- 7.8 23.2 +/- 6.5 -- U 1,277 +/- 40 4.4 +/- 0.1 5.2 +/- 0.1 5.9 +/- 0.1 U, total (ug/l) 2,640 9.1 10.6 12.1 NOTE: All units in pCi/l, unless indicated *U-238 concentrations were calculated. <MDA = less than the calculated Minimum Detectable Activity (MDA) for the counted sample Examples 13 and 14 demonstrate additional uses for the present invention. Sample material and RGW for Examples 13 and 14 were obtained from the Midwestern United States site. To establish baseline untreated characterization data, RGW and soil+RGW samples were tested for radionuclides and other radioactive substances using SPLP and RGW/TCLP extraction methods, prior to adding TGPA to the sample material. The following tests were performed: 1) RGW was tested for total radionuclides and other radioactive substances (US-7); PA1 2) RGW was mixed into the sample material at 30% (wt.). Radionuclides and other radioactive substances were examined in the amended sample material's SPLP extract (US-8); and PA1 (3) DI water was mixed into the sample material at 30% (wt.). Radionuclides and other radioactive substances were examined in the amended sample material's modified TCLP extract where RGW was utilized as the substitute TCLP extraction fluid (US-9). Table XIX presents the baseline data. Previous SPLP extraction test results from the same sample material amended only with DI water (US-5) are presented for comparison. TABLE XIX BASELINE RADIONUCLIDE LEACHABILITY FOR UNTREATED SAMPLE MATERIAL USING RADIOACTIVE GROUNDWATER (RGW) AS A DISPERSING AGENT AND EXTRACTION FLUID US-8 US-9 US-5 US-7 30% RGW 30% DI H.sub.2 O 30% DI Water Radionuclide/ RGW SPLP RGW as SPLP Isotope/Item Totals Extract TCLP Fluid Extract Bi-211 234 +/- 18 <MDA <MDA <MDA Ra-224 <MDA <MDA 254 +/- 131 <MDA Pb-212 <MDA <MDA 27.8 +/- 11.7 <MDA Ra-226 6 +/- 7 <MDA <MDA <MDA U-235 9,251 +/- 1,341 261 +/- 49 8,353 +/- 115 9,190 +/- 43 Th-234 35,940 +/- 5,027 560 +/- 113 26,220 +/- 462 3,355 +/- 90 U, total (mg/l) 97,431 7,813 66,471 16,375 U-238 (ug/l) 45,793 3,696 31,441 11,210 Tc-99 126,790 580 +/- 30 63,241 +/- 589 152 +/- 10 pH (S.U.) 7.5 TSS (mg/l) 1,320 TDS (mg/l) 4,400 Hardness [CaCO.sub.3 (mg/l)] 1,734 * U-238 concentrations were calculated. All units expressed as pCi/l, unless indicated <MDA = less than Minimum Detectable Activity for the counted sample. EXAMPLE 13 In Example 13, the effects of extracting TGPA treated radioactive sample material containing RGW with USEPA's simulated acid rain leaching method (SPLP) are presented. In this example, RGW was used as a dispersion agent in place of deionized water. Contaminated sample material (characterized in Table XIX) was mixed with RGW at 30% (wt.). Three (3) equivalent aliquots of the sample material mixed with RGW were placed in separate beakers. In the first beaker, TGPA was added at a dose of 2% (wt.) and mixed (TS-11). In the second beaker, TGPA was added at a dose of 5% (wt.) and mixed (TS-12). In the third beaker, TGPA was added at a dose of 10% (wt.) and mixed (TS-13). The amount of TGPA added was calculated from the base mass of the untreated sample material exclusive of the RGW mass added. Table XX presents the data from the analysis of SPLP extract for each of the treated samples (TS-11, 12, and 13). The untreated characterization data from samples (US-7, and US-8) are presented in Table XIX for comparison. The SPLP extraction (SW-846, Method 1312) is USEPA's procedure for simulating soil exposure to acid rain. The SPLP method calls for the extraction of 100 g of material with 2000 ml of simulated acid rain fluid. TABLE XX TGPA SOIL TREATMENT RESULTS: RADIONUCLIDES IN SPLP EXTRACT OF SAMPLE MATERIAL MIXED WITH 30% (WT.) RADIOACTIVE GROUNDWATER Radionuclide/ TS-11 TS-12 TS-13 Isotope/ Treated Treated Treated Item 2% TGPA 5% TGPA 10% TGPA Bi-211 <MDA <MDA <MDA Ra-226 <MDA <MDA <MDA U-235 <MDA <MDA <MDA Th-234 <MDA <MDA <MDA U, total, (mg/l) 30 19 38 U-238 (ug/l)* 14 9 18 Tc-99 292 .+-. 21 322 .+-. 23 280 .+-. 21 RGW (characterized in US-7) was added to TS-11, TS-12, and TS-13 at a dose of 30% (wt.) prior to TGPA addition. All units expressed as pCi/i, unless indicated. <MDA = less than Minimum Detectable Activity for the counted sample *U-238 concentrations were calculated. EXAMPLE 14 In Example 14, sample materials containing radionuclides and other radioactive substances was treated with varying doses of TGPA and DI water was utilized as a dispersing agent. These treated samples were then extracted using the modified TCLP method (RGW/TCLP) where RGW was substituted for the specified extraction fluid (TCLP Fluid No. 2). The sample material was mixed with DI water and three (3) equivalent aliquots of the material were placed in separate beakers. In the first beaker, TGPA was added at a dose of 2% (wt.) and mixed (TS-14). In the second beaker, TGPA was added at a dose of 5% (wt.) and mixed (TS-15). In the third beaker, TGPA was added at a dose of 10% (wt.) and mixed (TS-16). The percent weight of TGPA added was calculated from the initial base mass of the untreated sample material exclusive of the RGW mass added. Each of the treated samples were then extracted using the RGW/TCLP method with RGW fluid added at the method specified volume and ratio (100 g soil: 2000 ml fluid). Table XXI presents the data from the analysis of the modified RGW/TCLP extract for each of the treated samples (TS-14, 15, and 16). The untreated characterization data from RGW (US-7) and untreated soil extract by RGW/TCLP (US-9) are presented in Table XIX for comparison. TABLE XXI TGPA TREATMENT RESULTS: RADIONUCLIDES IN MODIFIED RGW/TCLP EXTRACT OF SAMPLE MATERIAL MIXED WITh 30% (WT.) DI WATER TS-14 TS-15 TS-16 Radionuclide/ 2% TGPA 5% TGPA 10% TGPA Isotope RGW/TCLP RGW/TCLP RGW/TCLP Bi-211 <MDA <MDA <MDA Ra-226 <MDA <MDA <MDA U-235 2,513 .+-. 461 1,919 .+-. 267 <MDA Th-234 <MDA 5,656 .+-. 790 200 .+-. 170 U, total (mg/l) 18,191 11,880 18 U-238 (ug/l)* 8,604 5,619 9 Tc-99 45,738 .+-. 222 60,398 .+-. 255 35,176 .+-. 195 2000 ml of RGW (characterized in US-7) was added as the TCLP extraction fluid to 100 g of the treated sample matrix. An units expressed as pCi/i, unless indicated. <MDA = less than Minimum Detectable Activity for the counted sample *U-238 concentrations were calculated. Examples 13 and 14 show that the present invention can use radioactive groundwater as a dispersing agent and that materials treated by the present invention can be used to treat RGW. These examples also demonstrate that acid rain will not affect treated material. EXAMPLE 15 Example 15 examines the leachability of constituents from a host material based on a calculation of the distribution coefficient (K.sub.d) for a given analyte (e.g., a specific constituent measured by the analyses). The distribution coefficient is expressed in ml/g and calculated as the quotient of the activity of nuclide sorbed per unit mass of host material (expressed in pCi/g), and the activity of the nuclide in extract solution per unit volume of extract (expressed in pCi/ml). K.sub.d is an equilibrium value often calculated to determine the sorption affinity of waste analytes (e.g., nuclides) by host matrix (e.g., contaminated material) in aqueous or other fluid suspensions. In this example, the distribution coefficients are calculated for the untreated (Table XXII) and TGPA treated material (Table XXIII). The same calculations can be made for similar extractions using other extraction fluids such as, deionized water, SPLP or RGW. TABLE XXII CALCULATED DISTRIBUTION COEFFICIENT (KD) OF UNTREATED SAMPLE MATERIAL MODIFIED USING SW-846, METHOD 1311 EXTRACTION METHOD US-10 US-1 US-1 Modified Total TCLP TCLP Distribution Activity Fluid 2 Fluid 2 Coefficient (K.sub.d) ANALYTE (pCi/g) (pCi/l) (pCi/ml) (ml/g) Pb-210 179 173 0.173 1,034.7 Bi-211 4,212 3,360 3.360 1,253.6 Bi-214 1,321 910 0.910 1,373.6 Fr-223 3,919 3,768 3.768 1,040.1 Ra-223 1,574 1,514 1.514 1,039.6 Ra-224 2,463 1,772 1.772 1,390.0 Ra-226 1,800 1,294 1.294 1,391.0 Ac-227 188 243 0.243 773.7 Th-227 960 921 0.921 1,042.3 Th-228 17,110 16,490 16.490 1,037.6 Pa-231 1,857 1,764 1.764 1,052.7 U-235 326 313 0.313 1,041.5 Th-234 NT 19 0.019 -- TABLE XXIII CALCULATED DISTRIBUTION COEFFICIENT (KD) OF TGPA TREATED SAMPLE MATERIAL MODIFIED USING SW-846, METHOD 1311 EXTRACTION METHOD US-10 TS-1 TS-1 Untreated 5% TGPA 5% TGPA Modified Material TCLP TCLP Distribution Total Activity Extract Extract Coefficient (K.sub.d) ANALYTE (pCi/g) (pCi/l) (pCi/ml) (ml/g) Pb-210 179 <82 <0.082 >2,183 Bi-211 4,212 <21 <0.021 >200,571 Bi-214 1,321 <21 <0.021 >62,905 Pb-214 1,250 <20 <0.020 >62,500 Fr-223 3,919 <226 <0.226 >17,341 Ra-223 1,574 <37 <0.037 >42,541 Ra-224 2,463 <50 <0.050 >49,260 Ra-226 1,800 <190 <0.190 >9,474 Ac-227 188 <44 <0.044 >4,273 Th-227 960 <56 <0.056 >17,143 Th-228 17,110 <588 <0.588 >29,099 Pa-231 1,857 <272 <0.272 >6,827 U-235 326 <104 <0.104 >3,135 Th-234 NT <12 <0.012 NA Tables XXII and XXIII show an increase of the sorption affinity of the radionuclides by the host material as a result of treatment with TGPA. Further, the calculations in Tables XXII and XXIII utilize the MDA values for the equation denominator. The MDA is based on numerous factors, including count times, background, detector efficiency, recovery, decay, and other variables. Therefore, the K.sub.d values for radionuclides in materials treated with TGPA are actually higher than what can be empirically determined when the nuclide presence in extract is <MDA. Although the present invention has been described in connection with preferred embodiments, it will be appreciated by those skilled in the art that additions, modifications, substitutions and deletions not specifically described may be made without departing from the spirit and scope of the invention defined in the appended claims. |
description | This application claims priority to U.S. Provisional Patent Application No. 62/441,038, filed on Dec. 30, 2016 and entitled: CONTROL ROD DASH POT INTEGRAL TO THE UPPER TIE PLATE, the contents of which are herein incorporated by reference in their entirety. This invention was made with Government support under Contract No. DE-NE0000633 awarded by the Department of Energy. The Government has certain rights in this invention. This disclosure generally relates to a control rod damping system. Dash pots constrict diameters near the bottom of the guide tubes slowing the fall of control rods during a scram to reduce potential impact damage. Low coolant flow through nuclear reactor guide tubes can lead to problems such as boiling, reduced fuel economy, and potential interference with control rode operations due to build-up of guide tube corrosion and precipitates. One potential cause of low coolant flow are the dash pots. A damping area or “dash pot” on the upper ends of control rods reduce the need to constrict the diameter of guide tubes. As a result, water can more freely flow through the guide tubes reducing boiling coolant issues. The restriction at the upper portion of the control rod assembly creates hydraulic back pressure which slows the fall and associated impact of the control rods hitting the fuel assembly during a scram procedure. The control rods include a first section having a first diameter retaining an active material for inserting into the guide tube and controlling a fission rate in a nuclear reactor core. A second section of the control rods attach to a head assembly. The novel dampening section is located between the first and second section with a second larger diameter. The dampening section reduces a separation distance between an outside surface of the control rod and an inside surface of the guide tube that decelerates the control rod when entering a top end of the guide tube. In one example, the control rod may have a cylindrical cladding including a bottom end retaining the active material and having a first wall thickness. A top end of the cladding may have a second continuously increasing wall thickness larger than the first wall thickness. In another example, the dampening area may be located on a drive shaft. The drive shaft may slidingly extend through an opening in a support member. The drive shaft may include a dampening section having a diameter larger than the opening in the support member to decelerate the drive shaft when dropped by a rod drive mechanism. FIG. 1 illustrates a cross-sectional view of an example reactor module 100 comprising reactor pressure vessel 52. Reactor core 6 is shown located near a lower head 55 of the reactor pressure vessel 52. The reactor core 6 may be located in a shroud 22 which surrounds reactor core 6 about its sides. A riser section 24 is located above the reactor core 6. When primary coolant 28 is heated by reactor core 6 as a result of fission events, primary coolant 28 may be directed from shroud 22 up into an annulus 23 located above reactor core 6, and out of riser 24. This may result in additional primary coolant 28 being drawn into shroud 22 to be heated in turn by reactor core 6, which draws yet more primary coolant 28 into shroud 22. The primary coolant 28 that emerges from riser 24 may be cooled down and directed towards the outside of the reactor pressure vessel 52 and then returned to the bottom of the reactor pressure vessel 52 through natural circulation. Primary coolant 28 circulates past the reactor core 6 to become high-temperature coolant TH and then continues up through the riser section 24 where it is directed back down the annulus and cooled off by a heat exchanger to become low-temperature coolant TC. One or more control rod drive mechanisms (CRDM) 10 operably coupled to a number of drive shafts 20 may be configured to interface with a plurality of control rod assemblies 82 located above reactor core 6. A reactor pressure vessel baffle plate 45 may be configured to direct the primary coolant 28 towards a lower end 55 of the reactor pressure vessel 52. A surface of the reactor pressure vessel baffle plate 45 may come into direct contact with and deflect the primary coolant 28 that exits the riser section 24. In some examples, the reactor pressure vessel baffle plate 45 may be made of stainless steel or other materials. The lower end 55 of the reactor pressure vessel 52 may comprise an ellipsoidal, domed, concave, or hemispherical portion 55A, wherein the ellipsoidal portion 55A directs the primary coolant 28 towards the reactor core 6. The ellipsoidal portion 55A may increase flow rate and promote natural circulation of the primary coolant through the reactor core 6. Further optimization of the coolant flow 28 may be obtained by modifying a radius of curvature of the reactor pressure vessel baffle plate 45 to eliminate/minimize boundary layer separation and stagnation regions. The reactor pressure vessel baffle plate 45 is illustrated as being located between the top of the riser section 24 and a pressurizer region 15. The pressurizer region 15 is shown as comprising one or more heaters and a spray nozzle configured to control a pressure, or maintain a steam dome, within an upper end 56 or head of the reactor pressure vessel 52. Primary coolant 28 located below the reactor pressure vessel baffle plate 45 may comprise relatively sub-cooled coolant TSUB, whereas primary coolant 28 in the pressurizer region 15 in the upper end 56 of the reactor pressure vessel 52 may comprise substantially saturated coolant TSAT. A fluid level of primary coolant 28 is shown as being above the reactor pressure vessel baffle plate 45, and within the pressurizer region 15, such that the entire volume between the reactor pressure vessel baffle plate 45 and the lower end 55 of the reactor pressure vessel 52 may be full of primary coolant 28 during normal operation of the reactor module 100. Shroud 22 may support one or more control rod guide tubes 124. The one or more control rod guide tubes 124 serve to guide control rod assemblies 82 that are inserted into, or removed from, reactor core 6. In some examples, control rod drive shafts 20 may pass through reactor pressure vessel baffle plate 45 and through riser section 24 in order to control the position of control rod assemblies 82 relative to reactor core 6. Reactor pressure vessel 52 may comprise a flange by which lower head 55 may be removably attached to a vessel body 60 of reactor pressure vessel 52. In some examples, when lower head 55 is separated from vessel body 60, such as during a refueling operation, riser section 24, baffle plate 45, and other internals may be retained within vessel body 60, whereas reactor core 6 may be retained within lower head 55. Additionally, vessel body 60 may be housed within a containment vessel 70. Any air or other gases that reside in a containment region 74 located between containment vessel 70 and reactor pressure vessel 52 may be removed or voided prior to or during reactor startup. The gases that are voided or evacuated from the containment region 74 may comprise non-condensable gases and/or condensable gases. Condensable gases may include steam that is vented into containment region 74. During an emergency operation, vapor and/or steam may be vented into containment region 74, only a negligible amount of non-condensable gas (such as hydrogen) may be vented or released into containment region 74. Certain gases may be considered non-condensable under operating pressures that are experienced within a nuclear reactor system. These non-condensable gases may include hydrogen and oxygen, for example. During an emergency operation, steam may chemically react with the fuel rods to produce a high level of hydrogen. When hydrogen mixes with air or oxygen, this may create a combustible mixture. By removing a substantial portion of the air or oxygen from containment vessel 54, the amount of hydrogen and oxygen that is allowed to mix may be minimized or eliminated. It may be possible to assume from a practical standpoint, that substantially no non-condensable gases are released into or otherwise housed in containment region 74 during operation of reactor module 100. Accordingly, in some examples, substantially no hydrogen gas is present in the containment region 74, such that the levels and/or amounts of hydrogen together with any oxygen that may exist within the containment region 74 are maintained at a non-combustible level. Additionally, this non-combustible level of oxygen-hydrogen mixture may be maintained without the use of hydrogen recombiners. In some examples, separate vent lines from the reactor pressure vessel 52 may be configured to remove non-condensable gases during start up, heat up, cool down, and/or shut down of the reactor. During the emergency scram condition, drive assemblies 10 may release drive shafts 20 dropping control rod assemblies 82 into guide tubes 124. Conventional guide tubes 124 may narrow toward bottom ends to hydraulically dampen the impact of control rod assemblies 82 dropping into reactor core 6. As described above, the narrow bottom diameters of guide tubes 124 may reduce the flow of primary coolant 28 through reactor core 6 causing coolant 28 to boil resulting in corrosion and reduced fuel economy. FIG. 2 is a perspective view of a control rod assembly 82 that includes dampening areas 130. Control rod assembly 82 may be held above and then inserted into reactor core 6. As explained above in FIG. 1, multiple drive shafts 20 extend from rod drive mechanisms 10, through baffle plate 45 and shroud 22 down to the top of reactor core 6. In one example, drive shafts 20 extend through a drive shaft support 122 that may be part of baffle plate 45 described above in FIG. 1. However, drive shaft support 80 may be located and attached anywhere within reactor pressure vessel 52, such on shroud 22, annulus 23 or riser section 24. A head assembly 86 may include a cylinder 88 that attaches to the bottom end of drive shaft 20. Head assembly 86 also may include arms 90 that extend radially out from cylinder 86 and attach at distal ends to top ends of control rods 92. Head assembly 86 is alternatively referred to a spider machining and control rods 92 are alternatively referred to as fingers. Control rods 92 extend into a fuel assembly 120 that is alternatively referred to as a fuel bundle and in FIG. 1 forms part of reactor core 6. Fuel assembly 120 may include a top nozzle 122 that supports multiple guide tubes 124. Guide tubes 124 extend down from nozzle 122 and in-between nuclear fuel rods (not shown). Control rods 92 control the fission rate of uranium and plutonium fuel rods. Control rods 92 are typically held by drive shaft 20 above fuel assembly 120 or held slightly inserted into fuel assembly 120. Reactor core 6 may overheat. A nuclear scram operation is initiated where rod drive mechanisms 10 in FIG. 1 release drive shafts 20 dropping control rods 92 down into guide tubes 124 and in-between the fuel rods. Some fuel assemblies narrow bottom ends of guide tubes 124 to reduce the impact of control rod assembly 82 slamming into fuel assembly 120. As explained above, these narrow diameters at the bottom ends of guide tubes restrict coolant flow causing steam created corrosion. Negative effects of low coolant flow can be even more detrimental in a nuclear reactor, such as nuclear reactor module 100 that may use natural circulation, instead of pumps, to circulate coolant through guide tubes 124. Dampening areas 130 are integrated into the upper ends of control rods 92 to reduce the impact of dropping control rod assembly 82 onto fuel assembly 120 during a scram operation. Instead of continuously restricting coolant flow through the bottom ends of guide tubes 124, dampening areas 130 only restrict coolant flow at the upper ends of guide tubes 124 during the scram operation. In addition, coolant flow is only restricted after control rods 92 are mostly inserted into guide tubes 124. In another example, dampening areas 150 are located on drive shafts 20 moving impact forces even further above control rod assembly 82 and fuel assembly 120. FIG. 3 is a side view and FIG. 4 is a side sectional view of control rod assembly 82 and fuel assembly 120. FIG. 5 is a more detailed side sectional view for a portion of control rod 92 that includes dampening area 130. Referring to FIGS. 3-5, guide tube sleeves 126 extend downward from substantially the middle of holes 128 formed in floor 123 of nozzle 122. Guide tubes 124 extend from a top surface of floor 123 through holes 128 and sleeves 126 down in between fuel rods of the reactor core. Control rods 92 each include a top plug section 136, an intermediate section 129 that holds a spring 134, and a bottom section 131 that holds active control rod material 132. Active material 132 is used in reactor core 6 of FIG. 1 to control the fission rate of uranium and plutonium. At least in some examples, active material 132 may include chemical elements such as boron, silver, indium and cadmium that are capable of absorbing neutrons without themselves fissioning. Each control rod 92 extends down from head assembly 86 into the top end of an associated guide tube 124. In a fully inserted position, control rods 92 extend through nozzle 122 and down to the bottom of guide tubes 124 in between the fuel rods. Control rods 92 are normally held by drive shaft 20 above nozzle 122 and are typically not completely inserted into fuel assembly 126 unless an overheating condition is detected. Dampening area 130 is located in the top ends of intermediate sections 129 between plug 136 and above active material 132 where spring 134 is located. As explained in more detail below, dampening area 130 reduces the impact when control rods 92 are dropped into guide tubes 124 during a nuclear scram. In one example, the diameters of control rods 92 in dampening area 130 are larger than the diameters of the lower sections 131 that extend down into fuel assembly 120. This allows substantially the entire lower section 131 carrying active material 132 to fully insert in between the fuel rods prior to dampening area 130 reaching the top ends of guide tubes 124. FIG. 6 is a top sectional view of control rod assembly 82 and guide tubes 124. FIG. 7 is a side sectional view of a control rod 92 partially inserted into an associated guide tube 124. FIG. 8 is a further enlarged detailed side sectional view of dampening area 130 formed in control rod 92. Referring first to FIGS. 4, 6 and 7, as explained above, plug 136 of control rod 92 includes a top end 137A that inserts into the bottom end of arm 90 on head assembly 86 and a bottom end 137B that inserts into a cylindrical cladding 140. In one example, cladding 140 has a circular cross-sectional shape that retains spring 134 and active material 132. In one example, cladding 140 may be made out of stainless steel. A guide tube collar 142 extends up from floor 123 of nozzle 122 as shown in FIG. 4. Guide tube 124 extends down from collar 142 through floor 123 of nozzle 122 and down to the bottom of fuel assembly 120. A sleeve collar 144 sits in hole 128 of nozzle 122 as shown in FIG. 4. Sleeve 126 extends down from collar 144 below floor 123 of nozzle 122. A spring 146 extends around the outside surface of guide tube 124 between collar 142 and collar 144. Referring now to FIG. 8, a wall thickness and an associated outside diameter of cladding 140 may continuously increase from a lower dampening location 130A to an upper dampening location 130B. This increased wall thickness and corresponding increased diameter reduces a spacing 148 between the outside surface of cladding 140 and an inside surface of guide tube 124. For example, space 148A between cladding 140 and guide tube 124 at dampening location 130A is larger than space 148B between the outside surface of cladding 140 and the inside surface of guide tube 124 at upper dampening location 130B. Referring to FIGS. 1-8, during normal operations, drive shaft 20 may hold control rods 92 almost completely above fuel assembly 120. During an overheating condition, rod drive mechanisms 10 in FIG. 1 release drive shafts 20 dropping control rod assembly 82. The lower sections 131 of controls rods 92 that contain active material 132 have a uniform smaller diameter and accordingly drop freely down into guide tubes 124. Control rods 92 may push coolant out the top and bottom ends of guide tubes 124. Control rods 92 continue to drop freely until bottom ends 130A of dampening area 130 reach the top ends of guide tubes 124. The continuously increasing diameter of dampening area 130 start reducing the spacing 148 at the top ends of guide tubes 124 between the outside surface of control rods 92 and the inside surfaces of guide tubes 124. Dampening area 130 starts restricting the coolant from escaping through the top ends of guide tubes 124. The restricted coolant creates a back hydraulic pressure that slows down and absorbs some of the energy from the control rods 92 falling inside of guides tubes 124. As a result, the coolant in guide tubes 124 acts like a hydraulic cylinder decelerating the falling speed of control rod assembly 82. One substantial advantage of using larger diameter dampening section 130 is that guide tubes 124 may remain at a consistent diameter throughout the entire length of fuel assembly 120. Thus, guides tubes 126 may avoid creating the boiling and corrosion problems that exist in guide tubes with narrow diameter bottom ends. Wider dampening areas 130 also may be easier to manufacture compared with changing a diameter at the bottom of guide tubes 124. Wider dampening areas 130 also may stiffen the upper ends of control rods 92 and reduce binding when control rods 92 are dropped into guide tubes 124. In one example, a bottom outside diameter 149A at damping location 130A may be around 9.677 millimeters (mms), lower spacing 148A may be around 0.866 mms, upper outside diameter 149B at dampening location 130B may be around 10.668 mms, upper spacing 148B may be around 0.375 mms, and the distance between lower dampening location 130A and upper dampening location 130B may be around 85 mms. The spacings, diameters, and distances of dampening area 130 may vary based on the size and weight of control rod assembly 82. The dimensions of dampening area 130 can also be varied to provide a more gradual deceleration of control rod assembly 82. For example, the length between lower dampening location 130A and upper dampening location 130B may be increased to provide a more gradual deceleration of control rod assembly 82. In another example, holes may be drilled through the top ends of guide tubes 124 to provide an alternative coolant escape path. In another example, cladding 140 may remain at a same uniform thickness. However, outside diameter 149A of cladding 140 still may continuously increase from lower dampening location 130A to upper dampening location 130B. For example, an extrusion processed used for forming cladding 140 may form a continuously increasing diameter within dampening area 130. In one example, plug 136 of control rod 92 shown in FIG. 7 may have substantially the same larger outside diameter 149B as the upper end of cladding 140. In another example, cladding 140 may maintain substantially the same diameter 149 and fully extend into guide tubes 124. Dampening area 130 may be formed in plug 136 and have a continuously increasing outside diameter starting from bottom end 137B and extending up to upper end 137A. The diameter at upper end 137A may be sized so arms 90 do not fall on top of nozzle 122 when control rods 92 are released during the scram. In yet another example, V-shaped slots may extend up from floor 123 of nozzle 122. The slots may receive arms 90 and decelerate and stop control rod assembly 82 before slamming into the top of fuel assembly 120. FIG. 9 is a side view of drive shaft 20 and control rod assembly 82 shown above in FIG. 2. FIG. 10 is a sectional view of a portion of drive shaft 20 and drive shaft support 80. Referring to FIGS. 9 and 10, drive shafts 20 may be used instead of control rods 92 to dampen the speed of control rod assembly 82 during a nuclear scram. A lower portion of drive shaft 20 may have a first outside diameter 152A. Lower dampening location 150A may start at first outside diameter 152A and continuously increase until reaching a second larger outside diameter 152B at upper dampening location 150B. Drive shaft 20 may maintain smaller outside diameter 152A below dampening location 150A and may maintain larger outside diameter 152B above upper dampening location 150B. In one example, the outside diameter of drive shaft 20 is increased by increasing a thickness 156 of drive shaft wall 154. Of course, the outside diameter 152 of drive shaft 20 also may be increased without increasing the thickness 156 of drive shaft wall 154 using known extrusion processes. Drive shaft 20 may have a cylindrical shape and dampening area 150 may have an inverted cone shape. A circular opening 158 in drive shaft support 80 may be formed with an inclining inverted cone shaped inside wall 160 that receives and retains dampening area 150. A diameter of opening 158 may continuously increase from a bottom side of support 80 to a top side of support 80. Drive shaft 20 below dampening location 150A can slide freely through opening 158 dropping control rod assembly 82 down into fuel assembly 120. The diameter at the bottom end of opening 158 is smaller than diameter 152B of drive shaft 20 at upper dampening location 150B. Accordingly, drive shaft 20 starts decelerating as the outside surface of dampening area 150 starts sitting against inside wall 160 of support 80. Dampening area 150 may stop drive shaft 20 before drive rod assembly 82 slams down against the top of nozzle 122. For example, dampening area 150 may stop drive rod 20 just before arms 90 of head assembly 86 reach nozzle 122 as shown in FIG. 5. Alternative dampening schemes may be used with drive shafts 20. For example, a spring may extend up from the top surface of support 80. A transverse bar or wider outside diameter 152B of drive rod 20 may compress the spring to decelerate and eventually stop drive rod 20. In another example, a cone shaped facet with upwardly inclining sides may extend up from the top surface of support 80 and operate similar to upwardly inclining wall 160 of support 80. In another example, dampening areas 130 in control rods 92 and dampening areas 150 in drive rods 20 may be used in combination to further distribute the impact of falling control rod assembly 82. Having described and illustrated the principles of a preferred embodiment, it should be apparent that the embodiments may be modified in arrangement and detail without departing from such principles. Claim is made to all modifications and variation coming within the spirit and scope of the following claims. Some of the operations described above may be implemented in software and other operations may be implemented in hardware. One or more of the operations, processes, or methods described herein may be performed by an apparatus, device, or system similar to those as described herein and with reference to the illustrated figures. It will be apparent to one skilled in the art that the disclosed implementations may be practiced without some or all of the specific details provided. In other instances, certain process or methods have not been described in detail in order to avoid unnecessarily obscuring the disclosed implementations. Other implementations and applications also are possible, and as such, the following examples should not be taken as definitive or limiting either in scope or setting. References have been made to accompanying drawings, which form a part of the description and in which are shown, by way of illustration, specific implementations. Although these disclosed implementations are described in sufficient detail to enable one skilled in the art to practice the implementations, it is to be understood that these examples are not limiting, such that other implementations may be used and changes may be made to the disclosed implementations without departing from their spirit and scope. Having described and illustrated the principles of a preferred embodiment, it should be apparent that the embodiments may be modified in arrangement and detail without departing from such principles. Claim is made to all modifications and variation coming within the spirit and scope of the following claims. |
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abstract | An apparatus for inspecting and testing a startup range neutron monitoring system for a nuclear reactor. The apparatus includes: a neutron-flux detector; a preamplifier that amplifies an electric signal output from the neutron-flux detector; a pulse measurement unit that counts times when electric signal output from the preamplifier exceeds a discrimination voltage; a discrimination-voltage setting unit that applies the discrimination voltage to the pulse measurement unit; a voltage-setting unit that applies a voltage to the neutron-flux detector; an arithmetic processing unit that calculates an output power of the reactor based upon an output signal of the pulse measurement unit; an output unit that outputs data representing the output power of the reactor, calculated by the arithmetic processing unit; and an inspecting/testing unit that sets the discrimination voltage and the voltage to be applied by the voltage-setting unit. |
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summary | ||
048266515 | abstract | Method and apparatus for assisting the loading of a reactor core with new and/or irradiated elongated fuel assemblies which includes inserting a fuel assembly with a fuel assembly carrier from a storage pit into a grid position of a reactor core with a refueling machine having propelling equipment, supplying an actual position of the fuel assembly carrier to the propelling equipment of the refueling machine during movement of the refueling machine between the storage pit and the reactor vessel, comparing the actual position of the fuel assembly carrier with a desired position of the fuel assembly carrier to find a deviation, and carrying out a correction of the deviation in the travelling movement of the refueling machine based upon the comparison. |
description | This application is a U.S. national phase under the provisions of 35 U.S.C. §371 of International Patent Application No. PCT/EP11/73482 filed Dec. 20, 2011, which in turn claims priority of French Patent Application No. 1060977 filed Dec. 21, 2010. The disclosures of such international patent application and French priority patent application are hereby incorporated herein by reference in their respective entireties, for all purposes. The present invention pertains to solid organic scintillators doped by one or more chemical elements comprising a specific polymeric matrix in which fluorophore compounds are dispersed, these scintillators having the combination of following characteristics: possible high doping level of chemical element(s); capability of emitting at long wavelengths (in particular at wavelengths of at least 550 nm); and optionally, a short scintillation decay time. The invention finds particular application in: the medical field, more particularly medical imaging using X-rays; the field of experimental physics, and more particularly the design of X-ray diagnostic imaging for Megajoule Laser plasma experiments. Conventionally a scintillator is intended for the detection of radiation or of high energy particles, which requires the use of material capable of using radiation-matter interaction to convert the energy of ionizing radiation derived from the particles to be detected (this ionizing radiation possibly being X-rays, γ rays for example whose energies range from a few keV to several MeV) into visible or near UV light. The emitted light is then analysed by a photodetector optionally coupled with the scintillator (this photodetector possibly being a photomultiplier, a photodiode, a photographic film or CCD camera). To be efficient, a scintillator must preferably meet the following characteristics: it must have good absorption efficacy of ionizing radiation; it must be able, for a given incident particle, to generate sufficient light so that it can be collected; it must be transparent to its own light to avoid losses; the light must be emitted sufficiently rapidly after passing of the particle, which in other words means that the scintillator must have a short scintillation decay time, in particular for X-ray imaging applications in a hostile radiating environment. Regarding radiation absorption capacity, this is related to the interaction mode between the matter and the rays, this interaction mode possibly being a photoelectric effect, Compton scattering effect and a pair production effect. Regarding X-rays, the absorption of these rays is related to the effective atomic number Zeff, this effective atomic number having to be the highest possible to allow complete absorption of the X-rays by the material via photoelectric effect for incident energies lower than 100 keV, complete absorption not being possible via Compton effect (arising through the radiation/matter interaction for materials having a low effective atomic number) in which only part of the energy is absorbed in the constituent material of the scintillator. Therefore the X-ray images obtained with materials which, after interaction with incident X photons generate a Compton effect, are of poor quality since the X photons diffusing in the material are not completely absorbed by the material or, if they are absorbed, may be absorbed in such manner that they contribute to obtaining a blurred image. At the current time, scintillators capable of allowing X-ray absorption via dominant photoelectric effect for incident energies lower than 100 keV are inorganic scintillators such as YAG-type scintillators, more particularly a scintillator of yttrium garnet and caesium doped aluminium type (Y3Al5O12:Ce). On the other hand, the solid organic scintillators conventionally used (typically having an effective atomic number Zeff in the order of 5) do not allow complete absorption of X-rays of energy ranging from 10 to 40 keV, which means that they are little useful for X-imaging. Regarding light emission, another phenomenon likely to distort X-ray dosimetry is so-called <<Cherenkov>> radiation, characterized by continuous emission intensity, energy loss being proportional to the inverse cube of the wavelength from the ultraviolet to blue-red. Finally so that they can be efficiently used in the field of X-ray imaging, it is desirable that scintillators should have a rapid mean scintillation decay time due to photon absorption, in particular so as to avoid degradation of X-ray imaging by highly energetic nuclear fusion products (such as fusion neutrons and γ rays, in particular for Megajoule Laser X-ray imaging). To meet these three conditions (namely, a scintillator having a high effective atomic number, a scintillator emitting fluorescence light of long wavelength and a scintillator advantageously having a short scintillation decay time), inorganic scintillators have been proposed but unconvincingly for X-ray imaging dedicated to Megajoule Laser experiments. For example: inorganic scintillators of YAG type are scarcely efficient having regard to their scintillation decay time that is too long (this decay time ranging from 60 to 100 ns); YI3:Ce, GdI3:Ce and LuI3 scintillators, while they deliver substantial fluorescent light under X-ray radiation, they nevertheless also have a long scintillation decay time (from 30 to 40 ns) and emit in the yellow. Additionally, most of these scintillators are hygroscopic, which limits their field of application, in particular for X-ray imaging. Faced with the difficulty of finding inorganic scintillators suiting their criteria (namely, capability of absorbing radiation essentially via photoelectric effect and capability of emitting intensely at long wavelengths, such as wavelengths of at least 550 nm), the authors of the present invention have focused on the development of solid organic scintillators meeting the above-mentioned criteria. The invention therefore relates to a solid organic scintillator comprising a polymeric matrix, in which one or more fluorophores compounds are dispersed together with one or more chemical elements having an atomic number of 40 to 83, characterized in that the said scintillator has a weight content of said chemical element(s) of at least 5% by weight relative to the total weight of the scintillator, and in that the scintillator emits an emission spectrum comprising an emission peak at a wavelength of at least 550 nm and advantageously has a scintillation decay time of less than 20 ns. With the above-mentioned characteristics, the scintillators conforming to the invention display numerous advantages. Through the significant presence of chemical element such as defined above translating as a high effective atomic number (Zeff) (e.g. at least 30) of the scintillator material, the scintillators of the invention have strong radiation-stopping power, in particular for low energy X-rays (e.g. ranging from 10 to 40 keV). In addition, again on account of the high atomic number (Zeff), the energy of the rays arriving on the scintillator (such as X-rays) of energy ranging from 10 to 40 keV is completely absorbed via photoelectric effect, which is of particular interest if the scintillator is intended to be used for X-ray imaging. Appended FIG. 1 illustrates this phenomenon. This Figure illustrates the three X-ray interaction modes (respectively photoelectric effect, Compton effect and dominant pair production) between X-rays having an incident energy in keV (shown as E along the abscissa) and a material having a given effective atomic number (this atomic number Zeff being shown along the ordinate). It can clearly be seen in this Figure that for a certain range of incident X-ray energy (from 0 to about 800 keV), the higher the effective atomic number (Zeff), the greater the probability of incident X-ray absorption via photoelectric effect. Also, since the scintillators of the invention are capable of emitting an emission spectrum having an emission peak at a wavelength of at least 550 nm (and possibly reaching 620 nm for example), it is possible to overcome a parasitic phenomenon usually encountered with scintillators emitting in the ultraviolet to blue: Cherenkov radiation, likely to distort results when the scintillators are used in the field of X-ray dosimetry. Further, the scintillators of the invention may have a scintillation decay time of less than 20 ns, preferably from 10 to 15 ns. As mentioned above, the scintillators of the invention comprise one or more chemical elements whose atomic number Z may range from 40 to 83. For example, the chemical elements can be chosen from among the elements having an atomic number ranging from 40 to 56 and from 72 to 83. In particular, the chemical elements can be chosen from among Zr (Z=40), Nb (Z=41), Mo (Z=42), Ru (Z=44), Rh (Z=45), Pd (Z=46), Ag (Z=47), Cd (Z=48), In (Z=49), Sn (Z=50), Te (Z=52), I (Z=53), Cs (Z=55), Ba (Z=56), La (Z=57), Ce (Z=58), Nd (Z=60), Sm (Z=62), Eu (Z=63), Gd (Z=64), Dy (Z=66), Tm (Z=69), Yb (Z=70), Ir (Z=77), Pt (Z=78), Au (Z=79), Tl (Z=81), Pb (Z=82) et Bi (Z=83). More particularly, the chemical elements able to be included as constituents of scintillators of the invention can be chosen from among Mo, Ag, Cd, In, Sn, I, Ba, Nd, Sm, Gd, Yb, Tl, Pb and Bi. Preferably, the chemical elements which may be included as constituents of the scintillators of the invention may be lead or tin. According to the invention, the chemical elements used to form the scintillators of the invention are present in a content of at least 5% relative to the total weight of the scintillator, in particular a content ranging from 10 to 27% by weight relative to the total weight of the scintillator. When incorporated at a content starting from 5 weight % relative to the total weight of the scintillator, a chemical element such as lead particularly allows a high atomic number (Zeff) to be imparted to the material of the scintillator as demonstrated by FIG. 2 which gives a graph of the trend in atomic number (Zeff) as a function of the lead weight content (Pb wt %). The chemical element weight content can be determined by elementary analysis. According to the invention, the polymeric matrix may comprise one or more (co)polymers, these (co)polymers advantageously imparting transparency properties to the matrix. In particular, the polymeric matrix may advantageously comprise a (co)polymer comprising repeat units resulting from polymerization of one or more monomers chosen from among styrene, vinyltoluene, vinylxylene, methyl methacrylate, methacrylic acid, 2-hydroxyethyl methacrylate. More particularly the polymeric matrix, according to a first embodiment, may comprise a copolymer comprising repeat units resulting from polymerization of vinyltoluene and methacrylic acid or, according to a second embodiment, a (co)polymer comprising repeat units resulting from polymerization of 2-hydroxyethyl methacrylate. The above-mentioned copolymer(s) may be cross-linked copolymers e.g. using a cross-linking agent which may be a monomer comprising at least two polymerizable functions capable, after polymerization, to form a bridge between two copolymer chains. As an example of cross-linking agent, mention can be made of dimethacrylate monomers. In this case, in addition to the above-mentioned repeat units, the copolymer will comprise repeat units derived from the polymerization of said cross-linking agent. The scintillator of the invention, as indicated above, comprises one or more fluorophore compounds. It is specified that by fluorophore compound is meant a chemical compound capable of emitting visible fluorescence light after excitation by photons or other incident particles. In our case, the fluorophore compound(s) advantageously have the characteristic of being able to absorb photons of wavelengths belonging to the UV region, and of re-emitting photons so that the emission spectrum of these compounds has an emission peak at a wavelength of at least 550 nm, which forms one of the characteristics of the scintillators as mentioned above. The scintillators of the invention may comprise a first fluorophore compound and a second fluorophore compound. Advantageously, the scintillators of the invention may comprise a first fluorophore compound capable of absorbing photons and, after this absorption, of emitting photons so that the emission spectrum of the first fluorophore compound has an emission peak at a wavelength of less than 550 nm, and a second fluorophore compound capable of absorbing the photons of wavelengths belonging to said emission spectrum of said first fluorophore compound and, after this absorption, of emitting photons so that the emission spectrum of said second fluorophore compound has an emission peak at a wavelength of at least 550 nm. A first fluorophore compound able to be incorporated in the scintillators of the invention may meet following formula (I): where: R1 is a mesomeric donor group; R2 is a hydrogen atom or mesomeric donor group the same or different from R1; R3 is: an acyl group; or a straight-chain or branched, saturated or unsaturated C1 to C20 hydrocarbon group, optionally substituted; or a cyclic, saturated C3 to C10 hydrocarbon group, optionally substituted; or a saturated C3 to C10 heterocyclic group optionally substituted; or an aryl or heteroaryl group, optionally substituted;or a salt thereof. According to the invention, the mesomeric donor group(s) represented by R1 and optionally R2, are preferably chosen from among: the —OR′ and —SR′ groups where R′ is a straight-chain or branched, saturated or unsaturated C1 to C20 hydrocarbon group, optionally substituted, or a C3 to C10 saturated cyclic hydrocarbon group, optionally substituted, or an optionally substituted aryl or heteroaryl group; and the —NR′R″ groups where R′ has the same meaning as previously, whilst R″ is either a hydrogen atom, or a straight-chain or branched, saturated or unsaturated C1 to C20 hydrocarbon group, optionally substituted, or an optionally substituted cyclic, saturated C3 to C10 hydrocarbon group, or an optionally substituted aryl or heteroaryl group. In the foregoing and in the remainder hereof by <<straight-chain or branched, saturated or unsaturated C1 to C20 hydrocarbon group>> is conventionally meant any alkyl, alkenyl or alkynyl group which comprises at least one carbon atom but no more than 20 carbon atoms. Said group may one of the following for example: methyl, ethyl, propyl, isopropyl, butyl, pentyl, neopentyl, hexyl, ethylenyl, propylenyl, butenyl, pentenyl, hexenyl, methylpentenyl, buta-1,3-dienyl, ethynyl, propynyl, butynyl, pentynyl, hexynyl, etc. By <<cyclic, saturated C3 to C10 hydrocarbon group>> is conventionally meant any group which is formed of a cycloalkyl or several fused cycloalkyls and which comprises at least 3 carbon atoms but no more than 10 carbon atoms. Said group may be one of the following for example: cyclopropyl, cyclobutyl, cyclo-pentyl, cyclohexyl, bicyclohexyl, bicyclodecyl, etc. By <<saturated C3 to C10 heterocyclic group>> is conventionally meant a monocyclic or polycyclic group containing one or more heteroatoms and which comprises at least 3 carbon atom but no more than carbon atoms. Said group may be one of the following for example: tetrahydrofuryl, tetrahydro-thiophenyl, pyrrolidinyl, piperidyl, dioxanyl, etc. By <<aryl group>> is conventionally meant a monocyclic or polycyclic group which meets Hückel's rule, i.e. whose number of delocalized electrons π is equal to 4n+2 (with n=0, 1, 2, 3, . . . ), and by <<heteroaryl group>> is meant a group such as just defined but which comprises one or more heteroatoms. As examples of an aryl group which may be used, mention can be made of the following groups: cyclopentadienyl, phenyl, benzyl, biphenyl, pyrenyl, naphthalenyl, phenantrenyl and anthrakenyl, whilst as examples of heteroaryl groups the following can be cited: furanyl, pyrrolyl, thiophenyl, oxazolyl, pyrazolyle, thiazolyl, imidazolyl, triazolyl, pyridinyl, pyranyl, quinolinyl, pyrazinyl and pyrimidinyl. Finally, by <<heteroatom>> is conventionally meant any atom other than carbon or hydrogen such as, for example, oxygen, sulphur, nitrogen, phosphorus atoms or a boron atom on the understanding however that the heteroatoms which may be ring members are oxygen, nitrogen or sulphur atoms. According to the invention, as cyclic, saturated C3 to C10 hydrocarbon groups and as saturated C3 to C10 heterocyclic groups, preferred use is made of monocyclic groups with 5 or 6 members. Similarly, as aryl or heteroaryl groups, it is preferred to use monocyclic groups with 5 or 6 members or polycyclic groups not containing more than 3 rings and, further preferably, no more than 2 rings each with 5 or 6 members. According to the invention, it is preferred to use derivatives meeting general formula (I) in which R1 is a —NR′R″ group, where R′ and R″ have the same meaning as given previously, whilst R2 represents a hydrogen atom. According to another preferred provision of the invention, R3 represents a relatively voluminous, sterically hindering group so as to limit the phenomenon of auto-quenching. It is therefore preferred that R3 represents a cyclic group, typically an aryl or heteroaryl group substituted one or more times by a branched C3 to C6 alkyl group such as an isopropyl or t-butyl group for example. A cyclic group of this type is the di-t-butylphenyl group for example. A first specific, particularly advantageous fluorophore compound is a fluorophore meeting following formula (II): The scintillators of the invention may comprise a proportion of first fluorophore compound ranging from 0.05 to 2% by weight, preferably 1 to 2% by weight relative to the total weight of the scintillator. The second fluorophore compound of the above-mentioned type may be a compound chosen from among perylene-diimide compounds, oxazone compounds, xanthene compounds, tetraphenylnaphthacene compounds, porphyrin compounds, pyrane compounds, triphenylmethane compounds and mixtures thereof. Regarding perylene-diimide compounds, mention can be made of compounds which meet following formula (III): where R4, R5, R6, R7, R8 and R9 independently of each other represent a hydrogen atom, a straight-chain or branched, saturated or unsaturated C1 to C20 hydrocarbon group, optionally substituted, an aryl or heteroaryl group optionally substituted. More particularly, R4 and R7 may represent a cyclic group, typically an aryl or heteroaryl group substituted one or more times by a branched C3 to C6 alkyl group, such as an isopropyl or t-butyl group for example. An example of a cyclic group of this type is the di-t-butylphenyl group and R5, R6, R8 and R9 may represent a hydrogen atom or an alkyl group, for example an alkyl group comprising 1 to 6 carbon atoms. A specific perylene-diimide compound meeting this definition meets following formula (IV): With regard to oxazone compounds, these may meet following formula (V): where R10, R11, R12, R13, R14, R15 and R16 independently of each other represent a hydrogen atom, a straight-chain or branched, saturated or unsaturated C1 to C20 hydrocarbon group, an —OH group, —NR′R″ group, R′ and R″ being such as defined above or, when positioned on two adjacent carbons, optionally an aromatic cyclic group provided that at least one of these groups represents an —OH group or —NR′R″ group. Specific compounds meeting this definition may meet one of following formulas (VI) and (VII): the compound of formula (VI) being known as Nile Red and the compound of formula (VII) being known as resorufine. In particular, Nile Red is characterized by a molar absorption coefficient ε of 34464 L·mol−1·cm−1 and a quantum yield in spectroscopic toluene of about 34% at 1.5*10−3 M and about 53% at 10−5 M. Regarding xanthene compounds, these can be more particularly chosen from among rhodamine compounds (such as rhodamine 6G, rhodamine B, rhodamine 123), eosin, fluorescein. Regarding tetraphenylnaphthacene compounds, particular mention can be made of rubrene (also known as 5,6,11,12-tetraphenylnaphthacene). Regarding pyrane compounds, 4-(dicyanomethylene)-2-methyl-6-(4-dimethylaminostyryl)-4H-pyran can be cited. Regarding porphyrins, 5,10,15,20-tetraphenylporphyrin can be cited. For triphenylmethane compounds, crystal violet can be mentioned (also known as methyl violet 10B). The scintillators of the invention may comprise a proportion of second fluorophore compound ranging from 0.002 to 1% by weight, preferably 0.05 to 1% by weight relative to the total weight of the scintillator. Particular scintillators conforming to the invention are scintillators chosen from among: scintillators in which: the chemical element is lead or tin; the polymeric matrix comprises a copolymer comprising repeat units derived from polymerization of vinyltoluene and methacrylic acid; the first fluorophore compound is a compound of formula (II) such as explained above and the second fluorophore compound is a compound of formula (IV) such as explained above; and scintillators in which: the chemical element is lead; the polymeric matrix comprises a (co)polymer comprising repeat units derived from polymerization of 2-hydroxyethyl methacrylate; the first fluorophore compound is a compound of formula (II) such as explained above and the second fluorophore compound is a compound of formula (IV) such as explained above. The scintillators of the invention can be characterized by one or more of the following characteristics: a capacity of absorbing photons of wavelengths possibly ranging from 350 to 500 nm; a capacity of emitting an emission spectrum having an emission peak at a wavelength of at least 550 nm and possibly of up to 620 nm; a scintillation decay time possibly ranging from 10 to 15 ns. From a morphological viewpoint, the scintillators of the invention may assume various forms, which are a function of the mould used. They may therefore be in the shape of cylinders of slabs. The scintillators of the invention, on account of their polymeric nature, can be prepared using different polymerization techniques allowing the concomitant incorporation of the above-mentioned chemical elements and above-mentioned fluorophore compounds. In particular, the scintillators of the invention can be prepared using a method comprising the following steps: a contacting step to place at least two monomers, of which at least one is complexed with at least one chemical element such as defined above, in contact with at least one fluorophore compound; a polymerization step of said monomers, after which the resulting product comprises a polymeric matrix in which said chemical element(s) and said fluorophore compound(s) are trapped. When the fluorophore compounds comprise at least one unsaturated hydrocarbon group (such as an ethylene group) said fluorophore compounds may directly take part in the polymerization reaction after which, on completion of the polymerization step, they will be covalently bonded to said matrix. The contacting step and the polymerization step can be conducted in a mould of size allowing the resulting product to have the desired dimensions of the scintillators of the invention. The monomers can be chosen from among styrene, vinyltoluene, vinylxylene, methyl methacrylate, methacrylic acid, 2-hydroxyethyl methacrylate. The monomers complexed with at least one chemical element may be monomers comprising at least two polymerizable functions, such as dimethacrylate monomers (e.g. lead dimethacrylate). The invention will now be described in connection with the Examples given below as non-limiting illustrations. Before going into more detail with the description of these Examples, first an explanation will be given of the protocols for measuring different magnitudes allowing characterization of the present invention, namely the weight content of chemical element (expressed as weight % relative to the total weight of the scintillator), the scintillation decay time (expressed in ns), the X absorption spectrum, the visible emission spectrum. Protocol for Measuring the Weight Content of Chemical Element The lead content is determined a priori by calculating the concentration of lead element relative to the other constituent elements of the scintillator (C, H, N, S, O) (this content is called the <<expected content>> in the Tables giving the results of the Examples. This lead content is then verified a posteriori by performing elementary analysis (this content being called the <<analysed content>> in the Tables giving the results of the Examples). For this purpose, a small portion of scintillator is cut off and ground. It is then analysed using apparatus capable of performing microanalyses. Protocol for Measuring Scintillation Decay Time The scintillation decay time of the scintillators of the invention is measured using a device illustrated in FIG. 3 comprising an electric arc discharge lamp in hydrogen gas 2, driven by a rack 1, a photomultiplier tube 7 and a single-photon counting line (composed of a rate counter 8, a time-amplitude converter 9, a delay box 10 and a pico-timing discriminator 11 and acquisition computer 12). The lamp 2 generates broad spectrum UV pulses of 1 ns duration, at a frequency of 40 Hz. These UV pulses are optically filtered with a high-pass interference filter 3 for best excitation of the tested scintillator 4 (as a function of a priori knowledge of the excitation spectrum of the solvent). The scintillator is placed in a closed support, insulating it from ambient light. The visible emission of the scintillator is then filtered with a band-pass filter 5 so as to isolate the most intense wavelength of the emission spectrum. An optical density 6 is placed between the scintillator and the photomultiplier tube. The anode of the photomultiplier tube 7 is connected to the pico-second discriminator 11. This discriminator 11 is connected to the rate counter 8. The rate indicated must not be higher than the driving frequency of the lamp 2 i.e. no more than one photon must be seen by the photomultiplier tube 7 per pulse of the lamp 2. If necessary, the intensity detected by the photomultiplier tube 7 is adjusted with the optical density 6. The rate counter 8 is connected to a time-amplitude converter 11, itself connected to the acquisition computer 12 (acting as oscilloscope), in which an acquisition card is housed. Acquisition lasts the time needed for acquiring a signal representing a decrease in light intensity of the scintillator 4 over at least four orders of magnitude. On completion of acquisition, a data file is collected in the form of a histogram associating a number of hits with channels, representing a relative intensity as a function of time. The curve is then analysed by means of a numerical spread-sheet. Starting from the moment of maximum intensity, the curve is adjusted using a decreasing exponential function. The so-called decay time is the time of the decreasing exponential function having the best coefficient of determination. Protocol for Determining the X Absorption Spectrum The determination of the X absorption spectrum of the scintillators of the invention is carried out using a device illustrated in appended FIG. 4 comprising an X-ray generator 15 and an X-ray detector in cadmium telluride 17 sensitive in the spectral region of 10 to 100 keV. The detector 17 is positioned 4.5 m away from the X-ray generator 15. The X-ray generator 15 generates an X spectrum derived from braking of electrons on a tungsten anode. The minimum detectable energy in the spectrum is 20 keV. The maximum energy of the X spectrum depends on the control voltage. For example, for a control voltage of 40 kV, the maximum energy contained in the spectrum is 40 keV and so forth for all possible voltages up to 160 kV. Radiation intensity is controlled by the intensity of the current of electrons striking the tungsten anode, and the exposure time of the scintillator 18. The output window of the generator 15 has a diameter size of 30 mm at 10 cm away from the source, which defines the cone 16 limiting the X-ray flow. The detector 17 is placed 4.5 m away from the output window of the generator 15. With this distance it is possible to obtain uniform irradiation of the sensitive surface of the detector 17 which measures 25 mm2. An X spectrum is recorded having parameters of 40 kV, 40 mA and 30 s, without any scintillator placed in front of the detector. This spectrum will be used as reference. This spectrum is pertinent for absorption measurements between 20 and 40 keV. The scintillator 18 is then laid against the radiation detector and the X spectrum is measured which meets the same parameters as recorded without scintillator. After collecting the histogram delivered by the detector, using a numerical spread-sheet the ratio is calculated between the total energy of the spectrum recorded with the scintillator and the total energy of the spectrum recorded without the scintillator. This calculation gives the scintillator's transmission of X-rays for a broad X spectrum. The complement to 1 of this ratio corresponds to the absorption of X-rays by the scintillator. Protocol for Determining the Visible Emission Spectrum Under X Excitation The visible emission spectrum of the scintillators of the invention is determined using a device comprising an X-ray generator tube with tungsten anode 19, a sample holder 22, a collecting optical fibre 23, a monochromator 24 and an acquisition computer 25 illustrated in appended FIG. 5. The scintillator is placed 21 to 30 cm away from the output shutter of the generator 19. The opening of the shutter is such that the X-ray beam 20 has a square section of 10 mm×10 mm on the surface of the scintillator 21. An optical fibre 23 is placed facing the scintillator 21 at a distance of 1 cm taking care that it does not lie in the axis of X-ray propagation emitted by the tube 21. The monochromator 24 comprises a diffraction grating of 300 lines per millimeter, 500 nm blaze wavelength. The luminescence originating from the scintillator 21 via the optical fibre 23, thus decomposed into its multiple components, is redirected by a set of mirrors onto a CCD camera having a sensitive silicon portion composed of 128 rows and 1024 columns. Each zone of the image reproduced by the CCD camera corresponds to a wavelength. Since the monochromator assembly is driven by computer, the luminescence spectrum of the scintillator 21 is reproduced in graph form showing relative intensity in number of hits along the Y-axis and the wavelength in nanometers along the X-axis. The scintillator conforming to the invention prepared according to this Example comprises a polymeric matrix formed of a poly(vinyltoluene-co-methacrylic acid) copolymer cross-linked with lead dimethacrylate, in which a first fluorophore compound is dispersed: N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide of following formula (II): and a second fluorophore compound: (bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide) of following formula (IV): Initially, preparation of the first fluorophore compound is performed after which the scintillator as such is prepared. The second fluorophore compound is commercially available. The title compound is prepared as per the following reaction scheme: Ac meaning acetyl and NMP meaning N-methylpyrrolidone. Compounds 1 and 2 are commercially available. In a 250 mL round-bottomed flask fitted with a water cooler, 2.934 g (10.06 mmol) of compound 1 and 4.131 g (20.12 mmol) of compound 2 are covered with 100 mL of freshly distilled quinoline. Next, 773 mg (3.52 mmol) of dihydrated zinc acetate are added and the reaction mixture is heated under reflux of the solvent for 5 hours. After return to ambient temperature, the mixture is poured into an aqueous solution of pH=1. The aqueous phase is extracted with dichloromethane. The organic phase is dried, filtered and concentrated. The residue is finally chromatographied on silica gel to give 4.583 g of beige solid (Yield: 93%). Melting point: 214° C. (dec., heptane) 1H NMR (250 MHz, CDCl3) δ ppm: 1.19 (s, 9H, CH3); 1.23 (s, 9H, CH3); 6.91 (d, 1H, J=2.2, H6′); 7.37 (dd, 1H, J=8.7, J=2.2, H4′); 7.48-7.52 (m, 1H, H3′); 7.80 (dd, 1H, J=8.5, J=7.2, H6); 7.99 (d, 1H, J=8.0); 8.38 (d, 1H, J=8.1); 8.55 (dd, 1H, J=8.5, J=1.3); 8.64 (dd, 1H, J=7.2, J=1.3) 13C NMR (62.9 MHz) δ ppm: 31.2, 31.7, 34.2, 35.4, 122.6, 123.5, 126.3, 127.6, 128.1, 128.7, 129.4, 130.5, 130.8, 131.1, 131.6, 132.4, 132.5, 133.5, 143.7, 150.1, 164.52, 164.57 Infrared (neat, cm−1): 2960, 2873, 1666, 1589, 1496, 1357, 1234. Synthesis of the Title Compound In a 10 mL round-bottomed flask, N-(2′,5′-di-t-butylphenyl)-4-bromo-1,8-naphthalimide (411 mg, 0.88 mmol) is placed in suspension in N-methylpyrrolidone (abbreviated to NMP). The addition is made of n-butylamine (437 μL, 4.42 mmol) and the reaction mixture is heated to 80° C. for 24 hours. After return to ambient temperature, the reaction mixture is purified by passing through a silica gel column (previously prepared by the addition of powder silica to dichloromethane). After re-crystallization in ethyl ether, the N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide is isolated in the form of a yellow powder (403 mg, 98%). The product obtained has a melting point of 233° C. (in diethyl ether). b) Preparation of the Scintillator as Such Pure methacrylic acid (3 g, 34.8 mmol), pure vinyltoluene (5 g, 42.3 mmol), lead dimethacrylate (2 g, 5.3 mmol), N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide (5 mg, 0.011 mmol) and bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide (0.2 mg, 0.26 mmol) are mixed in an inert atmosphere in a dry flask. The mixture is freed of any gas using the <<freeze-pump-thaw>> method and is then carefully poured into a mould which will impart the final shape to the scintillator. The mixture placed in the mould is subjected to a heat cycle (first at 65° C. for 4 days; secondly at 70° C. for 2 days and thirdly at 100° C. for 1 day). The resulting product is released from the mould and polished until a surface condition is obtained that is optically compatible with imaging applications. It is in the form of an orange-coloured cylinder of diameter 12 mm and height of 4 mm. The Table below summarizes the weight percentages (expressed relative to the total weight of the mixture) of the ingredients used to prepare this scintillator. IngredientsWeight %Vinyltoluene50Methacrylic acid30Lead dimethacrylate20First fluorophore compound0.05Second fluorophore compound0.002 For the scintillator obtained, the following measurements were made: the weight content of lead by elementary analysis; measurement of mass density; determination of the absorption spectrum obtained using a spectrofluorimeter, observing fluorescence intensity at different excitation wavelengths; determination of its emission spectrum by exciting the molecule to maximum excitation then using a spectrofluorimeter to record the fluorescence intensity obtained; the scintillation decay time, which is obtained by exciting the scintillator with a UV flash lamp; the effective atomic number, determined using XμDAT software from the exact proportion of each element measured by microanalysis. The results are grouped together in the following Table. MagnitudeResultsDensity (in g · cm−3)1.17Absorption spectrum (in nm)350-500Maximum peak of the emission579spectrum (in nm)Decay time (in ns ±1 ns)13.3Effective atomic number40.3Lead weight contentExpected: 11%Analysed: 12.3% The scintillator conforming to the invention prepared in this Example comprises a polymeric matrix formed of a poly(2-hydroxyethyl methacrylate) polymer cross-linked with lead dimethacrylate, in which are dispersed a first fluorophore compound: N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide of following formula (II): and a second fluorophore compound: (bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide) of following formula: The first fluorophore compound is prepared following the operating protocol described in Example 1. The second fluorophore compound is commercially available. The scintillator as such is prepared as per the following operating protocol. Pure 2-hydroxyethyl methacrylate (5 g, 38.4 mmol), lead dimethacrylate (5 g, 13.2 mmol), N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide (5 mg, 0.011 mmol) and bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide (0.2 mg, 0.26 μmol) are mixed in an inert atmosphere in a dry flask. The mixture is freed of any gas using the <<freeze-pump-thaw>> method and carefully poured into a mould which will impart the final shape to the scintillator. The mixture placed in the mould is subjected to a heat cycle (first at 85° C. for 24 hours, secondly at 120° C. for 5 hours and thirdly at 90° C. for 48 hours). The resulting product is released from the mould and polished to obtain a surface condition optically compatible with imaging applications. The scintillator is obtained in the form of an orange-coloured cylinder 42 mm in diameter and 3 mm in height. The Table below summarizes the weight percentages (expressed relative to the total weight of the mixture) of the ingredients used to prepare this scintillator. IngredientsWeight %2-hydroxyethyl methacrylate50Lead dimethacrylate50First fluorophore compound0.05Second fluorophore compound0.002 For the scintillator obtained, the following measurements were determined: weight content of lead; density measurement by calculating the ratio: scintillator weight/volume; determination of the absorption spectrum obtained, using a spectrofluorimeter and observing the fluorescence intensity at different excitation wavelengths; determination of its emission spectrum by exciting the molecule to maximum excitation and using a spectrofluorimeter to record the fluorescence intensity obtained; the scintillation decay time; the effective atomic number determined using XμDAT software from the exact proportion of each element measured by microanalysis. The results are grouped together in the following Table. MagnitudeResultsDensity (in g · cm−3)1.55Absorption spectrum (in nm)350-500Maximum peak of the emission591spectrum (in nm)Decay time (in ns ±1 ns)9.2Effective atomic number53.1Weight content of leadExpected: 27.4%Analysed: 29.5% The scintillator conforming to the invention prepared in accordance with this Example comprises a polymeric matrix comprising a poly(vinyltoluene-co-methacrylic acid) copolymer cross-linked with n-dibutyltin dimethacrylate in which are dispersed a first fluorophore compound: N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide of following formula (II): and a second fluorophore compound: (bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide) of following formula (IV): Initially the first fluorophore compound is prepared followed by the preparation of the scintillator as such. The second fluorophore compound is commercially available. The scintillator as such is prepared as per the following operating protocol. Pure methacrylic acid (2.525 g, 29.3 mmol), pure vinyltoluene (2.525 g, 21.4 mmol), n-dibutyltin dimethacrylate (3.367 g, 8.3 mmol), N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide (4.2 mg, 0.01 mmol) and bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide (0.17 mg, 0.02 μmol) are mixed in an inert atmosphere in a dry flask. The mixture is freed of any gas using the <<freeze-pump-thaw>> method and carefully poured into a mould which will impart the final shape to the scintillator. The mixture placed in the mould is heated to 85° C. for 13 days. The resulting product is gently returned to ambient temperature, released from the mould and polished to obtain a surface condition optically compatible with imaging applications. It is in the form of an orange-coloured cylinder 42 mm in diameter and 2.3 mm in height. The scintillator conforming to the invention prepared following this Example comprises a polymeric matrix comprising a poly(vinyltoluene-co-methacrylic acid) copolymer cross-linked with lead dimethacrylate in which are dispersed a first fluorophore compound: N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide of following formula (II): a second fluorophore compound: (bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide) of following formula (IV): and a third fluorophore compound: 2,5-diphenyloxazole. The scintillator as such is prepared as per the following operating protocol. Pure methacrylic acid (5.38 g, 62.5 mmol), pure vinyltoluene (8.93 g, 75.6 mmol), lead dimethacrylate (3.584 g, 9.5 mmol), N-(2′,5′-di-t-butylphenyl)-4-butylamino-1,8-naphthalimide (17.9 mg, 0.040 mmol), bis-N-(2,5-di-t-butylphenyl)-3,4,9,10-perylenetetracarbodiimide (1.8 mg, 2.35 μmol) and 2,5-diphenyloxazole (358 mg, 1.62 mmol) are mixed in an inert atmosphere in a dry flask. The mixture is freed of any gas using the <<freeze-pump-thaw>> method and carefully poured into a mould which will impart the final shape to the scintillator. The mixture placed in the mould is heated to 65° C. for 4 days. The resulting product is gently returned to ambient temperature, released from the mould and polished to obtain a surface condition optically compatible with imaging applications. It is in the form of an orange-coloured cylinder of diameter 47 mm and height of 4.8 mm. |
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description | 1. Field of the Invention The present invention relates to a radiation treatment system designed to treat a tumor or the like in the body or the body surface of a patient by irradiating the tumor with particle beams such as a proton beam, a carbon beam or the like. More particularly, the invention relates to a radiation treatment system and a method thereof, capable of maintaining flatness, which indicates the degree of uniform radiation beam irradiation in an area to be exposed to radiation, by dividing the area to be exposed to radiation including its peripheral area into a plurality of areas, and then performing the simulation of radiation exposure according to the shape of each area, and improving the efficiency of radiation utilization. 2. Description of the Related Art The radiotherapy is a method of treatment employed to reduce or even remove a tumor by intensively applying radiation to the tumor generated in the body of a patient. When such a radiation exposure treatment is carried out, the area to be exposed to radiation including the tumor to be treated must be uniformly irradiated with a proper dose of radiation. On the other hand, radiation applied to a healthy organization around the tumor outside the area to be exposed to radiation must be suppressed as much as possible. Now, description will be made as to the radiation treatment system which uses particle beams among the radiations used for the above-described radiotherapy. Here, to briefly explain the principle of the radiation treatment system, consideration is given to a case where an area to be treated is a ball having a radius r, a patient is regarded as a cube having one side set equal to 4r, and the center of the ball as the area to be treated is located at the center of the cube. FIG. 9 illustrates the configuration of a radiation treatment system using such particle beams available in the related art. In the drawing, a reference numeral 101 denotes a region to be treated by radiotherapy (hereinafter may be referred to just as treatment region), in which a tumor or the like to be subjected to radiation exposure is present. This region is assumed to be a ball having a radius r, and the center of the ball is located at a cubic center of a patient 102 assumed as a cube having one side set equal to 4r. A reference numeral 102 denotes a patient fixed to a treatment couch 114, and irradiated with particle beams; 103 a radiation exposure region defined in a three-dimensional area which includes a treatment region 101; 104 a treatment planning device designed to perform treatment simulation for forming a radiation exposure region 103 according to the state of the diseased part of the patient 102 to be treated, and setting parameters for the direction of irradiation, the position of irradiation, and so on; 105 an accelerator controller for controlling the operation of an accelerator 107, specifically designed to adjust a radiation beam to appropriate strength according to the acceleration condition in the treatment simulation performed by the treatment planning device 104; and 106 an irradiation controller designed to control a wobbler device 108, a scatterer device 109, a dose monitoring device 110, a ridge filter device 111, a range shifter device 112, and a collimator device 113 respectively according to a set condition in the treatment simulation, thereby controlling the irradiation direction, position, and so on, of the radiation beam. A reference numeral 107 denotes an accelerator for offering energy to the radiation beam, a cyclotron, a synchrotron, or the like for accelerating the radiation beam containing charged particles by the acceleration electric field of high frequency being used therefor; 108 a wobbler device used to expand the radiation beam corresponding to the radiation exposure region 103, composed of a deflection electromagnet, and adapted to move the radiation beam so as to draw a circular orbit on the radiation exposure region 103 by applying a sinewave current having a phase different by 90° to this deflection electromagnet; 109 a scatterer device composed of a scatterer for scattering the radiation beam, and used to expand the radiation beam with the wobbler device 108 corresponding to the radiation exposure region 103; 110 a dose monitoring device for monitoring the dose of the radiation beam applied to the radiation exposure region 103, designed to output each dose of the radiation beam monitored to the irradiation controller 106; 111 a ridge filter device used for modulating the range of the radiation beam, generally made of brass or the like having a proper shape on the surface, and designed to adjust the expanse in the advancing direction of the radiation beam; 112 a range shifter device for adjusting a reaching distance in the advancing direction of the radiation beam according to the set condition in the treatment simulation; 113 a collimator device for adjusting the passage aperture of the radiation beam corresponding to the radiation exposure region 103; and 114 a treatment couch for laying the patient 102 thereon. Next, the operation of the system as configured above will now be described as below. First, before a particle beams treatment is carried out, the image data of a diseased part obtained by photographing the diseased part (equivalent to the treatment region 101) of the patient 102 with an X-ray CT not-shown is output to the treatment planning device 104. Based on the state of the diseased part analyzed from the input image data thereof, the treatment planning device 104 decides a radiation exposure region 103 by adding an area or the like as a margin to the treatment region 101, and then performs treatment simulation for setting parameters for the direction of irradiation, the position of irradiation, and so on. In this case, in the radiation treatment system available in the related art, when the treatment region 101 is formed to be a ball having a radius r, a radiation exposure region 103 becomes a circle having a radius r. Then, the ridge filter device 111 having a width of 2r of SOBP (spread out Bragg peak) in the advancing direction of the radiation beam is used. Moreover, in the accelerator 107, the radiation beam is accelerated by acceleration energy that causes the reaching distance of particle beams in the body of the patient 102 to be 3r. Then, the treatment simulation is executed by the wobbling device 108 and the scatterer device 109 such that the radiation beam is uniformly applied in the radiation exposure region 103 (flattening). Besides the foregoing, in order to accurately irradiate the radiation exposure region 103 with the radiation beam, there are cases where a collimator dedicated to the patient having a cylindrical radiation beam passage aperture is used, and the collimator device 113 having general applicability is used. FIG. 9 shows the example of using the collimator device 113. Now, the treatment simulation will be described in detail. The treatment planning device 104 calculates a parameter to be set for the collimator device 113, assuming that the collimator is usually circumscribed on the circular radiation exposure region 103 having a radius r. Then, corresponding to the radiation exposure region 103, the treatment planning device 104 selects operation conditions for the wobbler device 108 and the scatterer device 109 for expanding the radiation beam. In addition, the ridge filter for causing the expanse in the advancing direction of the radiation beam to be 2r is selected. In this case, if the reaching distance of the radiation beam is not exactly 3r in the body of the patient 102, then the treatment planning device 104 selects an operation condition for the accelerator 107 for offering radiation beam energy to realize a reaching distance of 3r or more. Subsequently, a difference in the reaching distance equivalent to the acceleration energy offered by the accelerator 107 and the reaching distance 3r of the radiation beam in the body of the patient 102 is adjusted by the range shifter device 112. The treatment planning device 104 decides an irradiation condition of the radiation beam to coincide with the deepest part in the radiation exposure region 103, and also decides the irradiation dose of particle beams at this time. Checking is made as to the appropriateness of the dose distribution decided by the treatment planning device 104. If appropriate, then the treatment planning device 104 outputs the set parameter of the radiation treatment system which has been obtained in the treatment simulation, to the accelerator controller 105 and the irradiation controller 106. Upon having received the treatment parameter, the accelerator controller 105 and the irradiation controller 106 set the above-mentioned treatment parameter in each of the accelerator 107, the wobbler device 108, the scatterer device 109, the dose monitoring device 110, the ridge filter device 111, the range shifter device 112, and the collimator device 113. Then, the patient 101 is laid on the treatment couch 114 and fixed, and the radiation exposure region 103 is aligned with the position of irradiation, and irradiated with particle beams. When the dose monitoring device 110 determines that the prescribed dose of radiation beam has been applied, the irradiation with the radiation beam is stopped, completing one treatment. Next, description will be made as to a general method for performing flattening to uniformly irradiate the radiation exposure region 103 with the radiation beam, used in the particle-beams treatment. Hereinafter, the irradiation of the radiation exposure region 103 uniformly with the radiation beam is simply referred to as the flattening of a radiation field for the easiness of explanation. In general, for the flattening of the radiation field, there are available a double scatterer system using a double scatterer, and a wobbler system using the wobbler device. These systems are both designed to form a circular radiation field (circular radiation exposure region 103). More specifically, the double scatterer system increases the level of scattering the radiation beam by using two kinds of scatterers arranged away from each other in the axial direction of the radiation beam, and flattens the radiation field by performing scattering in such a way as to increase the efficiency of using the radiation beam. There is a close relation between the size of the radiation field (radiation exposure region 103), and the scattering conditions of the two kinds of scatterers and the shapes thereof. The efficiency of using particle beams in the double scatterer system is generally about 30%. The efficiency of using particle-beams is equivalent to the ratio of a dose of radiation applied in the radiation field with respect to all doses of radiation including a dose of radiation applied outside the radiation field for the flattening of the radiation field. The wobbler system rotates the radiation beam by the wobbler device for generating a rotating magnetic field, scatters the radiation beam by the scatterer to expand a beam diameter, and then forms a radiation field having prescribed flatness (degree of uniformity when the radiation exposure region 103 is irradiated with the radiation beam, given by a difference in the reaching amounts of the radiation beams applied to the radiation exposure region 103) when the radiation beam makes one rotation. In this wobbler system, when the radiation field is enlarged, the rotational radius of the radiation beam is increased, and the thickness of the scatterer for expanding the radiation beam is increased. There is a relation given by an equation below among the size (rmax) of the radiation field, the rotational diameter (R0) of the radiation beam, and the expanse (σa) of the radiation beam, when the flatness of the radiation field is ±2%. In this case, the radiation field is formed based on the characteristic of the radiation beam supplied from the accelerator according to a prescribed relational equation. In the wobbler system, since particle beams in the area of 84% inside the rotational radius of the radiation beam is used, the efficiency of using particle beams becomes about 30%.R0:σa:rmax=1.00:0.90:0.84Here, rmax in the above-mentioned relational equation becomes small when the flatness is further increased to reach ±1%, and the efficiency of using the particle beams is lowered. The radiation treatment system of the related art is constructed in the foregoing manner, and the treatment simulation is performed for the circular radiation exposure region 103 formed by the double scatterer system or the wobbler system to decide parameters to be set for the respective devices. Consequently, the efficiency of using particle beams has been low. To explain the foregoing problem more specifically, an actual radiation exposure region 103 is not always a circular radiation field. Thus, when the radiation field is subjected to flattening by the double scatterer system or the wobbler system, the proportion of particle beams applied outside the radiation exposure region 103 is increased, bringing about a reduction in the efficiency of using particle beams. For example, the efficiency of particle beams in the foregoing wobbler system was theoretically 30%. Actually, however, the efficiency is lower than this value, and if the radiation exposure region 103 occupies only ½ of the circular radiation field, then the efficiency of using particle beams is lowered to 15%. The present invention was made to solve the foregoing problems. Objects of the invention are to provide a radiation treatment system and a radiation treatment method, capable of maintaining flatness, which is a degree of uniformly irradiating a region to be exposed to radiation with a radiation beam, and increasing the efficiency of using particle beams, by dividing a radiation exposure region including a region to be irradiated with particle beams and a peripheral region thereof into a plurality of unit radiation exposure regions, and then executing radiation treatment simulation according to the shape of each divided region. In accordance with an aspect of the invention, there is provided a radiation treatment system, comprising: simulation means for executing radiation treatment simulation for dividing a radiation exposure region and a peripheral region thereof to be irradiated with particle beams into a plurality of unit radiation exposure regions, and then applying particle beams according to a shape of each divided unit radiation exposure region; and radiation treatment planning means for obtaining a radiation treatment condition for causing flatness, which is a degree of uniformly irradiating the radiation exposure region with a proper dose of particle beams, to be in a desired range, and a dose of particle beams applied to the unit radiation exposure region of the peripheral region to be minimized, in the case where the simulation means executes the radiation treatment simulation, and then making a radiation treatment plan reflecting the radiation treatment condition. According to the radiation treatment system of the invention, the simulation means divides the radiation exposure region and the peripheral region thereof into unit radiation exposure regions of grid forms. According to the radiation treatment system of the invention, the simulation means divides the radiation exposure region and the peripheral region thereof into belt-like unit radiation exposure regions. According to the radiation treatment system of the invention, the simulation means divides the radiation exposure region and the peripheral region thereof into concentric circular unit radiation exposure regions. According to the radiation treatment system of the invention, when the unit radiation exposure region is located in a boundary of the radiation exposure region, the radiation treatment planning means determines a degree of contribution made by a dose of particle beams applied to the unit radiation exposure region located in the boundary to the radiation exposure region, according to a dose of particle beams applied to the unit radiation exposure region of the peripheral region. In accordance with another aspect of the invention, there is provided a radiation treatment method, comprising: a simulation step for dividing a radiation exposure region and a peripheral region thereof to be irradiated with particle beams into a plurality of unit radiation exposure regions, and then executing radiation treatment simulation according to a shape of each divided unit radiation exposure region; a radiation treatment planning step for obtaining a radiation treatment condition for causing flatness, which is a degree of uniformly irradiating the radiation exposure region with a proper dose of particle beams, to be in a desired range in the case where the simulation step is executed and a dose of particle beams applied to the unit radiation exposure region of the peripheral region to be minimized, and then making a radiation treatment plan reflecting the radiation treatment condition; and a radiation exposure step for applying particle beams to the radiation exposure region and the peripheral region thereof to be irradiated according to the radiation treatment plan made in the radiation treatment planing step. According to the radiation treatment method of the invention, in the simulation step, the radiation exposure region and the peripheral region thereof are divided into unit radiation exposure regions of grid forms. According to the radiation treatment method of the invention, in the simulation step, the radiation exposure region and the peripheral region thereof are divided into belt-like unit radiation exposure regions. According to the radiation treatment method of the invention, in the simulation step, the radiation exposure region and the peripheral region thereof are divided into concentric circular unit radiation exposure regions. According to the radiation treatment method of the invention, in the radiation treatment planning step, when the unit radiation exposure region is located in a boundary of the radiation exposure region, determination is made as to a degree of contribution made by a dose of particle beams applied to the unit radiation exposure region located in the boundary to the radiation exposure region, according to a dose of particle beams applied to the unit radiation exposure region of the peripheral region. Next, the preferred embodiments of the present invention will be described. FIG. 1 illustrates the configuration of a radiation treatment system according to the first embodiment of the invention. In the drawing, a reference numeral 1 denotes a treatment region, in which a tumor or the like is present to be subjected to radiation treatment, which treatment region being assumed to be a ball having a radius r, and the center thereof being located at a cubic center of a patient 2 assumed to be a cube having one side set to 4r; 2 a patient fixed to a treatment couch 14, and subjected to radiation exposure; 3 a radiation exposure region to be irradiated with particle beams, defined by a three-dimensional region including the treatment region 1; and 4 a treatment planning device (simulation means and irradiation planning means), adapted to form a radiation exposure region 3 according to the state of the diseased part of the patient 2, divide the thus formed radiation exposure region 3 and the peripheral region thereof in a grid form, and set a control parameter (radiation treatment condition) for each device to apply particle beams according to a grip point (unit irradiation region) in an accelerator controller 5 and an irradiation controller 6. Besides the foregoing, the treatment planning device 4 obtains a control parameter for each device to cause the flatness of the radiation exposure region 3 to be in a desired range, and the number of grid points irradiated with particle beams in the peripheral region to be minimum, when the accelerator controller 5 and the irradiation controller 6 execute treatment simulation (radiation treatment simulation) according to the above-described set control parameter, and then makes a radiation treatment plan reflecting this. A reference numeral 5 denotes an accelerator controller (simulation means) for controlling the operation of a controller 7, adapted to adjust radiation beam strength in the treatment simulation according to the control parameter input from the treatment planning device 4; 6 an irradiation controller (simulation means), adapted to execute the treatment simulation by controlling each of a wobbler device 8, a scatterer device 9, a dose monitoring device 10, a ridge filter device 11, a range shifter device 12, and a collimator device 13 according to a set condition of the control parameter input from the treatment planning device 4; 7 an accelerator for offering energy to the radiation beam, for which a cyclotron or a synchrotron for accelerating the radiation beam composed of charged particles by the acceleration electric field of high frequency as in the case described above in the related art; 8 a wobbler device for irradiating the grid point with the radiation beam by providing a prescribed angle of deflection to particle beams with a deflection electromagnet, an energizing current for which is controlled by irradiation controller 6; and 9 a scatterer device for receiving the entry of the radiation beam output from the wobbler device 8, and providing a prescribed shape to the radiation beam, composed of a scatterer for scattering the radiation beam. A reference numeral 10 denotes a dose monitoring device for monitoring a dose of the radiation beam applied to the radiation exposure region 3, adapted to output each monitored dose of the radiation beam to the irradiation controller 6; 11 a ridge filter device used for modulating the range of the radiation beam, made of brass or the like having a proper shape formed in the surface as in the case described above in the related art, and designed to adjust the expanse in the advancing direction of the radiation beam; 12 a range shifter device for adjusting the reaching distance of the radiation beam in its advancing direction according to the set condition in the treatment simulation; 13 a collimator device for adjusting the passage aperture of the radiation beam corresponding to the radiation exposure region 3; and 14 a treatment couch for laying the patient 2. Now, description will be made as to the radiation treatment system using particle beams according to the first embodiment. FIG. 2 shows the example of division into a radiation exposure region and a peripheral region thereof according to the first embodiment. In FIG. 2, the radiation exposure region 3 is divided into a plurality of squares having one side a in each of X and Y axes, with a treatment center equivalent to the center of the radiation exposure region 3 as an original point. In addition, the peripheral region outside the radiation exposure region 3 is similarly divided. The radiation treatment system of the first embodiment applies particle beams to the intersection (grid point) of longitudinal and horizontal lines constituting a boundary line of each square having one side a in FIG. 2. Description will be made as to a flattening condition for the radiation exposure region 3 (referred to as the flattening of a radiation field, hereinafter), before the explanation of the operation of the radiation treatment system of the first embodiment. Generally, the beam of particle beams can be approximated to Gaussian distribution. In the first embodiment, the radiation beam is processed as one subjected to Gaussian distribution. First, the flattening condition of the radiation field when particle beams having two-dimensional Gaussian distribution are used is obtained as follows. When an isotropic beam shape of two-dimensional Gaussian distribution is equivalent to standard deviation σxy, the beam shape (beam size) of particle beams having two-dimensional Gaussian distribution is given by an equation (1) described below. In addition, when the radiation beam is moved in a plane including regions divided at equal intervals a (stepsize ΔX and ΔY), and the respective points are irradiated by equal doses, a dose distribution can be calculated by an equation (2) decried below: d ( x i , y j , x , y ) = 1 2 πσ x y 2 e - ( x - x i ) 2 + ( y - y j ) 2 2 σ x y 2 ( 1 ) D ( x , y ) = ∑ i , j d ( x , y , x i , y j ) ( 2 ) Here, xi and yj are given by an equation (3) below:xi=i×Δx, yj=j×Δy (3)In this case, i, j≡±1, ±2, and is ±3, and ±4 . . . FIG. 3 is a graph showing a relation between a value of a stepsize of a radiation beam standardized by an isotropic beam shape of particle beams (standard deviation σxy) and the flatness of the radiation field. As shown in the same graph, assuming the relation of Δx=Δy, from the foregoing relational equation, the flatness of the radiation field is gently and monotonously increased with respect to values of the stepsizes ΔX, and ΔY of the radiation beam standardized by the beam shape (standard deviation σxy). Here, for example, assuming that the flatness of the radiation field is ±1%, then it can be understood that a stepsize is to be set equal to the beam shape (standard deviation σxy) by 1.8 times. This means that if a division interval a is set to an optionally-given beam shape by 1.8 times, the flatness of the radiation field in a horizontal direction becomes ±1%. Accordingly, if a stepsize in the horizontal direction has been decided, then the beam shape only needs to be set to a/1.8. Assuming that flatness is ±2%, as can be understood from FIG. 3, a stepsize only needs to be set equal to the beam size by about 1.9 times. On the other hand, to set the flatness of the radiation field within the range of ±1% as described above, a stepsize of particle beams only needs to be set equal to the beam size (standard deviation σxy) by 1.8 times or less than 1.8 times. This means that when particle beams are applied in a grid form, a contribution is made from particle beams applied from the most adjacent grid point away by 1.8σxy, but since a next most adjacent grid point away by 1.8σxy is totally separated by 3.6σxy, the contribution of particle beams can be almost ignored. In other words, at a grid point outside the end of the radiation field (peripheral region of the radiation exposure region 3), there is no need of irradiating the grid point away by 5.4σxy with particle beams in terms of safety. Therefore, the flatness of the radiation field can be sufficiently secured by performing radiation exposure from the end of the radiation field (boundary of the radiation exposure region 3) to the grid point away by 3.6σxy. As for the condition that the desired flatness of the radiation field like that described above is obtained, and the grid point in the peripheral region of the radiation exposure region 3 minimized (that is, dose of particle beams applied to the peripheral region is minimized), a specific control parameter is calculated by the treatment planning device 4 during the execution of the treatment simulation. Next, the operation of the system as configured above will now be described as below. First, before the execution of particle-beams treatment, the diseased part (equivalent to the treatment region 1) of the patient 2 is photographed by an X-ray CT, not shown, and then the obtained image data of the diseased part is output to the treatment planning device 4. Based on the state of the diseased part analyzed from the input image data of the diseased part, the treatment planning device 4 decides a radiation exposure region 3 by adding a region or the like as a margin to the treatment region 1. In this case, in the treatment simulation, in which a region to be irradiated with particle beams is formed at the treatment planning device 4 according to the first embodiment, the radiation exposure region 3 and the peripheral region thereof are divided in grid forms as shown in FIG. 2, and all the grid points thereof (unit radiation exposure regions) are irradiated with equal doses of particle beams. The treatment planning device 4 calculates a control parameter for irradiating each of the above grid points with particle beams, and outputs the control parameter to the accelerator controller 5 and the irradiation controller 6. According to the control parameter from the treatment planning device 4, the irradiation controller 6 sets a leaf control parameter for the collimator device 13 to be circumscribed on the radiation exposure region 3 which is assumed to be a circle having a radius r. In addition, the irradiation controller 6 controls the operations of the wobbler device 8, the scatterer device 9, the dose monitoring device 10, the ridge filter device 11, and the range shifter device 12. Then, the accelerator controller 5 associatively controls the accelerator 7. Accordingly, particle beams having a beam size decided by the interval length of the grid points shown in FIG. 3 are generated, and treatment simulation is executed (simulation step). Here, to irradiate the grid points with particle beams, the energizing current of the wobbler device 8 is controlled by the irradiation controller 6. Then, the irradiation can be executed by providing a prescribed angle of deflection to the particle beams. In the treatment simulation executed in the foregoing manner, determination is made as to up to which of the grid points located outside the end of the radiation exposure region 3 is to be irradiated with particle beams by the treatment planning device 4, in order to obtain desired flatness inside the radiation exposure region 3. In addition, in the foregoing case, no particle beams need to be applied to a portion away by 3 grid points or more from the end of the radiation exposure region 3. Thus, a coordinate of the grid point to be irradiated with particle beams can be calculated by the treatment simulation. Now, the reason for the unnecessity of applying particle beams to the portion away by 3 grid points or more from the end of the radiation exposure region 3 will be described. Strength is reduced to 0.368(1/e;e=2.71828 . . . ) of strength of the center portion, when separation is made by only a distance equivalent to the standard deviation σ in two-dimensional Gaussian distribution, from the center of the radiation beam (beam axis). Because strength is generally reduced to less than 0.01 of strength of the center portion when separated by 3σ, particles entering the radiation exposure region 3 from the grid point away by 3σ from the center portion gives almost no influence on the flatness (0.01) of the radiation field of 1%, which is considered to be less than 0.01. Accordingly, in the first embodiment, contribution made from the grid point away from 3 grid points or more from the end of the radiation exposure region 3 to the inside of the same is ignored. In the treatment simulation executed in the foregoing manner, according to a relation like that shown in FIG. 3, the treatment planning device 4 calculates a control parameter for providing a radiation treatment condition, by which a desired flatness of a radiation field is set, and a dose of particle beams applied to the outside of the radiation exposure region 3 is minimized. Then, by using the dose monitoring device 10, determination is made as to whether the dose distribution of particle beams applied by the treatment planning device 4 according to the control parameter is appropriate or not. If the appropriate dose distribution is determined, then, the treatment planning device 4 makes a radiation treatment plan reflecting the control parameter for the grid point to be irradiated with particle beams (radiation treatment planning step). The control parameter in the radiation treatment plan is output from the treatment planning device 4 to the accelerator controller 5 and the irradiation controller 6. The accelerator controller 5 and the irradiation controller 6 set the same control parameter as that of the treatment simulation input from the treatment planning device 4 in the accelerator 7, the wobbler device 8, the scatterer device 9, the dose monitoring device 10, the ridge filter device 11, and the range shifter device 12. In addition, open-degree data corresponding to the radiation exposure region 3 is calculated, and set in the collimator device 13. Then, the patient 2 is laid on the treatment couch 14 and fixed, and the radiation exposure region 3 is aligned with a position for radiation exposure. The irradiation controller 6 controls the wobbler device 8 to set particle beams in a position corresponding to the grid point to be irradiated with particle beams, when the radiation exposure region 3 is subjected to radiation exposure. In addition, the irradiation controller 6 sets a scatterer for providing a prescribed beam shape to the scatterer device 9, and then sets a dose of particle beams to be applied in the dose monitoring device 10 according to the radiation treatment plan. After the control parameter has been set in the foregoing manner, the irradiation of each grid point with particle beams is started. When a dose monitored by the dose monitoring device 10 reaches a prescribed value, then the irradiation controller 6 stops the radiation beam and performs control in such a way as to move particle beams to a next grid point. Then, the same dose of particle beams is set on the dose monitoring device 10, and the radiation exposure is carried on (radiation exposure step). After the foregoing operation has been carried out for all the grid points to be irradiated with particle beams, one radiation treatment is completed. According to the first embodiment, if no particle beams are applied to the peripheral region located within 3 grid points from the end (boundary) of the radiation exposure region 3 to flatten the radiation field, flatness inside the radiation exposure region 3 cannot be secured. Consequently, a part of the particle beams applied to the peripheral region located within 2 grid points from the end (boundary) of the radiation exposure region 3 is wasted. In other words, assuming that the radiation exposure region 3 is a circle having a radius r and its diameter 2r is equivalent to the number n of grid points, the number of grid points necessary for flattening the radiation field can be approximated by a circle area having a diameter (n+4) obtained by adding an amount equivalent to two grid points from the end of the radiation exposure region 3. The efficiency of using particle beams is given by the ratio of the number of grid points in the radiation exposure region 3 and the number of grid points in a region where the 2 grid points of the peripheral region are added to the radiation exposure region 3. Accordingly, the efficiency is given by (n/(n+4))2 considering that particle beams outside the radiation exposure region 3 are all wasted. Therefore, it can be understood that to secure 30% efficiency of using particle beams, n must be set equal to 5 or higher. In addition, the radiation exposure region 3 was a circle in the foregoing. Generally, however, the radiation exposure region 3 is not a circular radiation field. Thus, the efficiency of using particle beams was lower than the theoretical value of 30% in the method of the related art. For example, if the radiation exposure region 3 occupies only ½ of the circular radiation field, then the efficiency of using particle beams is lowered to 15%. On the other hand, according to the first embodiment, particle beams are applied only to the necessary grid points outside the actual radiation exposure region 3. Thus, even if the radiation exposure region 3 is divided into less than 5 in a grid form in the directions of X and Y axes, the actual efficiency of using particle beams can be increased. In addition, in the foregoing description, it is shown that the radiation exposure region 3 and the peripheral region thereof are divided from the treatment center in the directions of X and Y axes as shown in FIG. 2. However, a similar advantage can be obtained even by setting grid points in an orthogonal directions with a given point used as a reference. Further, the division was made into squares having one side a in FIG. 2. However, another division is possible. FIG. 4 shows another example of division for the radiation exposure region and the peripheral region thereof according to the first embodiment. As shown, the radiation exposure region 3 and the peripheral region thereof may be divided into squares having one side of a/2, stepsizes in the directions of X and Y axes may be set to a, and radiation exposure may be carried out by shifting a/2 when a next line is irradiated. In this way, the flattening condition of the radiation field can be obtained in the treatment simulation as in the case of the first embodiment. Thus, a similar advantage can be achieved. In addition, the radiation exposure region 3 and the peripheral region thereof were divided into the squares. However, even if theses regions are divided into given identical shapes r (diamond, or rectangle) other than squares, a dose of particle beams applied to each grid point for flattening the radiation exposure region 3 is reverse-calculated with the dose of particle beams applied to each grid point set as a parameter. In this case, if there is a solution to the calculation, then a similar advantage can be obtained. In addition, the analysis was made assuming that the beam shape (standard deviation σxy) isotropic as shown in FIG. 3. However, even the beam shape is not isotropic, simulation for optimizing a stepsize according to the beam shape can be carried out. Thus, a similar advantage can be obtained. In addition, the case where the radiation exposure region 3 was a circle was described. However, even if the radiation exposure region 3 takes a given shape, a similar advantage can be obtained by carrying out simulation for applying particle beams to a region away by 2 grid points from the end of the radiation exposure region 3. Further, the case where the flatness of the radiation exposure region 3 was ±1% was described. However, even if flatness is set to a given value, a similar advantage can be obtained. As apparent from FIG. 3, generally, the stepsize is smaller as the flatness of the radiation exposure region is improved. The stepsize is larger as the flatness of the radiation exposure region 3 is degraded. The dose of particle beams applied to the grid point located outside from the end of the radiation exposure region 3 is increased by increasing the flatness of the radiation exposure region 3. However, the dose of particle beams applied to the grid point located inside the radiation exposure region 3 is also increased. On the other hand, if the flatness is lowered, the dose of particle beams applied to the grid point located outside from the end of the radiation exposure region 3 is reduced. However, the dose of particle beams applied to the grid point located inside the radiation exposure region 3 is also reduced. Considering the above relation, it can be understood that the efficiency of using particle beams is not so dependent on the simulated flatness if the range of flatness to be used for radiation exposure treatment is about 0.5–5%, and the standard deviation σxy of the beam shape is about 1.7–2.1. In addition, in the example described above with reference to FIG. 2, there was no mention of the order of irradiating the grid points with particle beams. However, since the irradiation order of the grid points is not so important for the formation of the radiation exposure region 3, a similar advantage can be obtained even in the order of the grid points facilitating the control of the wobbler device 8 or in the order facilitating control data creation. Moreover, if the contribution of the dose of particle beams in each grid point is the same, a similar advantage can be provided even if irradiation is carried out not just once but by a plurality of times. In addition, the example of division made such that the grid point coincided with the end of the diameter part of the circular radiation exposure region 3 was shown. However, even if the grid point does not coincide with the end of the diameter part of the radiation exposure region 3, a similar advantage can be obtained by simulating the dose of particle beams applied to the grid point located outside the radiation exposure region 3 and including the grid point contributing to the radiation exposure region 3. Further, description was made of the example of irradiating the grid points with the similar doses of particle beams having similar beam shapes when the grid points were regularly disposed. However, though calculation becomes a complex equation, a similar advantage can be obtained if there is solution to a reverse-problem for flattening the inside of the radiation exposure region 3, even when particle beams having different beam shapes are applied to irregularly disposed grid points. In addition, assuming that the exposure position to be irradiated with particle beams changes timewise, if there is reproducibility or regularity in such a change, for the actual exposure position of each grid point, it is also possible to calculate a dose of particle beams applied to each grid point by the method described above. On the other hand, if there is no reproducibility or regularity in the change, then, a flattening condition for the radiation exposure region 3 can be obtained by including the width of the change in the beam shape of the applied particle beams. As described above, according to the first embodiment, the radiation exposure region 3 and the peripheral region thereof to be irradiated with particle beams are divided in the grid forms, and the treatment simulation for applying particle beams according to each divided grid point is carried out. During the treatment simulation, the radiation treatment condition for causing the flatness of the radiation exposure region 3 to be in a desired range, and the dose of particle beams applied to the grid point of the peripheral region to be minimized is obtained, the radiation treatment plan is made reflecting this radiation treatment condition and, based on this treatment plan, the radiation exposure region 3 and the peripheral region to be irradiated are subjected to radiation exposure. Thus, since the unit radiation exposure regions obtained by dividing the radiation exposure region 3 and the peripheral region thereof into pluralities of regions are used, it is possible to accurately decide a radiation exposure region for providing desired flatness, and to increase the efficiency of using particle beams compared with the case of the related art. Moreover, since particle beams are applied to the minimum grid points in the peripheral region for providing desired flatness, it is possible to suppress the generation of superfluous particle beams. The radiation treatment system of the second embodiment is basically similar in configuration to that of the first embodiment. Thus, description will be made of portions different from those of the first embodiment. First, the treatment planning device 4 (simulation means and irradiation planning means) of the second embodiment forms a radiation exposure region 3 according to the state of the diseased part of the patient 2 to be treated as in the case of the first embodiment, divides the radiation exposure region 3 and the peripheral region thereof into belt forms, and then sets a control parameter (radiation treatment condition) of each device for applying particle beams according to the center Line of the belt form (unit radiation exposure region) in the accelerator controller 5 and the irradiation controller 6. The accelerator controller 5 and the irradiation controller 6 execute treatment simulation for applying particle beams according to the center Line of the belt regions divided based on the control parameter. In the treatment simulation, the treatment planning device 4 obtains a control parameter of each device for causing the flatness of the radiation exposure region 3 to be in a desired range and the dose of particle beams to be applied to the peripheral region to be minimized, and then makes a radiation treatment plan reflecting the control parameter. Also, in the second embodiment, as in the case of the first embodiment, a treatment region 1 is assumed to be a ball having a radius r, and the center thereof is located at the cubit center of the patient 2 assumed to be a cube having one side 4r. FIG. 5 shows the example of dividing the radiation exposure region in the radiation treatment system of the second embodiment. As shown, division is made such that a width a is set in parallel with an X axis from the center line of the belt region to be irradiated with particle beams, and a division range is expanded to the outside of the radiation exposure region 3. The boundary of division is given by y=ma(m=±1/2,±3/2,±5/2, . . . ). The center line for applying particle beams is given by y=ka(k=0,±1, ±2,±3, . . . ) in the broken line of FIG. 5. Now, before the explanation of the operation of the radiation treatment system of the second embodiment, description will be made as to a condition for flattening the radiation exposure region 3 (referred to as the flattening of the radiation field, hereinafter). According to the second embodiment, it is assumed that particle beams are moved to irradiate the exposure belt-like regions with equal doses of particle beams per unit time, and the treatment planning device 4 evaluates flatness inside the radiation exposure region 3 by using the equation described above with reference to the first embodiment. However, since the equation (3) is for the movement of particle beams on a straight line, an equation (4) below is used:Yj−kabut k=0, ±1, ±2, and ±3 . . . (4) In this case, similarly to the first embodiment, If it is assumed that the beam shape of particle beams is subjected to two-dimensional Gaussian distribution, then, by obtaining the integration of the equation (1) under the condition of the equation (4), a dose distribution when the equal dose of particle beams is applied along the center line can be calculated. FIG. 6 is a graph showing a relation between a value of a belt division width a standardized with a beam shape (standard deviation σxy) and flatness according to the second embodiment (Δx=Δy). As shown, it can be understood that the radiation exposure region 3 only needs to be divided in belt forms by the pitch of 1.9σxy in order to set the flatness of ±1%. FIG. 7 shows a dose distribution (amount of particle beams) in the section in the dividing direction of the radiation exposure region according to the second embodiment. A 1.9 line pitch is set when the flatness is ±1% as shown in FIG. 6. Thus, as shown in FIG. 7, when a beam shape (beam size) (standard deviation σxy) is 10 mm, the division width a of the radiation exposure region 3 becomes 1.9σxy, and thus the positions of X=19, 38, and an integral multiple of 19 are irradiated with particle beams. In other words, if a beam shape σxy is 10 mm, particle beams are scanned in the direction of X by 19 mm. In addition, in the example shown in FIG. 7, prescribed flatness is obtained by X=38 (2σxy) to 57 (3σxy). The same applies to the center line, on which particle beams are moved. In this case, when particle beams are moved on the center line, a distance for irradiating the outside of the radiation exposure region 3 with superfluous particle beams is set equal to the width of nearly two belts. This is attributed to the fact that it is impossible to obtain desired flatness inside the radiation exposure region 3 only by a dose of particle beams from the end of the radiation exposure region 3, provided from particle beams moved on the center line of the belt of the outermost side inside the radiation exposure region 3. In other words, desired flatness is obtained by contribution of particle beams moved on the center line of the outside nearest from the end of the radiation exposure region 3 (1.9σxy) and contribution of particle beams moved on the center line located farther outside from the end of the radiation exposure region 3 (3.8σxy). On the other hand, contribution (5.7σxy) of particle beams moved on the center line of the belt located farther outside from the end of the radiation exposure region 3 becomes a size to be ignored. Here, the irradiation with the same dose of particle beams per unit time means a condition for causing the speed of moving particle beams on the center lines of the respective belts to be equal to one another. Accordingly, loads placed on the electromagnet system for generating a magnetic field on Y axis side to move particle beams are also equal. As for the condition for obtaining the foregoing desired flatness of the radiation field, and the number of belts in the peripheral region of the radiation exposure region 3 to be minimized (the dose of particle beams applied to the peripheral region is minimized), a specific control parameter is calculated therefor by the treatment planning device 4 during the treatment simulation. Next, the operation of the system as configured above will now be described as below. First, before the execution of particle-beam treatment, the diseased part (equivalent to the treatment region 1) of the patient 2 is photographed by the X-ray CT, not shown. Then, the obtained image data of the diseased part is output to the treatment planning device 4. Based on the state of the diseased part analyzed from the input image data of the diseased part, the treatment planning device 4 decides a radiation exposure region 3 by adding a region as a margin to the treatment region 1. In this case, it is assumed that in the treatment simulation carried out by the treatment planning device 4 to form a region to be irradiated with particle beams according to the second embodiment, as shown in FIG. 5, the radiation exposure region 3 and the peripheral region thereof are divided in belt forms, and an equal dose of particle beams is applied on the center line of each belt (unit radiation exposure region). The treatment planning device 4 calculates a control parameter for applying particle beams on the center line of the belt, and outputs the control parameter to the accelerator controller 5 and the irradiation controller 6. According to the control parameter from the treatment planning device 4, the irradiation controller 6 sets leaf-control parameters for the collimator device 13 to be circumscribed on the radiation exposure region 3 assumed to be a circle having a radius r, and then controls the operations of the wobbler device 8, the scatterer device 9, the dose monitoring device 10, the ridge filter device 11, and the range shifter device 12. Associatively, the accelerator controller 5 controls the accelerator 7, generates particle beams of a beam shape decided by the interval of the belts (unit radiation exposure regions) shown in FIG. 5, and then executes treatment simulation (simulation step). Here, to apply particle beams in a belt form, the energizing current of the wobbler device 8 is controlled to generate a magnetic field for providing a deflection angle in the direction of a Y axis, equivalent to the position of the belt, and generate a magnetic field moved at a fixed speed in the direction of an X axis for moving particle beams. In the treatment simulation carried out in the foregoing manner, at the treatment planning device 4, determination is made as to up to which of the belts located outside from the end of the radiation exposure region 3 are to be irradiated with particle beams, in order to obtain desired flatness inside the radiation exposure region 3. In addition, in the foregoing case, no particle beams need to be applied to portions away by 3 belts or more from the end of the end of the radiation exposure region 3. Accordingly, a coordinate of the belt to be irradiated with particle beams can be calculated by the treatment simulation. Then, the treatment planning device 4 calculates the dose of particle beams to be applied according to a distance of movement on the center line of each belt. Subsequently, by using the dose monitoring device 10, the treatment planning device 4 determines whether a dose distribution of particle beams applied according to the control parameter is appropriate or not. If the appropriate dose distribution is determined, then, the treatment planning device 4 makes a radiation treatment plan reflecting the control parameter of the belt to be irradiated with particle beams (radiation treatment planning step). The control parameter of the radiation treatment plan is output from the treatment planning device 4 to the accelerator controller 5 and the irradiation controller 6. The accelerator controller 5 and the irradiation controller 6 set the same control parameter as that for the treatment simulation input from the treatment planning device 4 in the accelerator 7, the wobbler device 8, the scatterer device 9, the dose monitoring device 10, the ridge filter device 11, and the range shifter device 12. Further, leaf-control data corresponding to the radiation exposure region 3 is calculated, and set in the collimator device 13. Then, the patient 2 is laid on the treatment couch 14 and fixed, and the radiation exposure region 3 is aligned with the position to be irradiated. When particle beams are applied to the radiation exposure region 3, the irradiation controller 6 controls the wobbler device 8 to set particle beams in the position corresponding to the belt to be irradiated with particle beams, sets a scatterer for providing a prescribed beam shape in the scatterer device 9, and then sets a dose of radiation rays to be applied in the dose monitoring device 10 according to the radiation treatment plan. After the foregoing setting of the control parameter, the application of particle beams to each belt is started. When a dose monitored by the dose monitoring device 10 reaches a prescribed value (particle beams are moved by a prescribed distance), the irradiation controller 6 stops the radiation beam and performs control so as to move the particle beams to a next belt. Then, a prescribed dose is set in the dose monitoring device 10, and the irradiation with particle beams is continued (radiation exposure step). After the foregoing operation has been executed for all the belts to be irradiated with particle beams, one radiation exposure treatment is completed. According to the second embodiment, for the flattening of the radiation field, If no particle beams are applied to the peripheral region of three belts or less from the end (boundary) of the radiation exposure region 3, flatness inside the radiation exposure region 3 cannot be secured. Consequently, a part of particle beams moved on the center line of the upper and lower two belts from the end (boundary) of the radiation exposure region 3 is wasted. In other words, assuming that the radiation exposure region 3 is a circle having a radius r, and its diameter 2r is equivalent to n belts, it is possible to approximate the number of belts necessary for flattening the radiation field by the area of a circle having a diameter (n+4), formed by adding two belts from the end of the radiation exposure region 3. The efficiency of using particle beams is given by the ratio of the number of belts in the radiation exposure region 3 and the number of belts in the region, to which the two belts of the peripheral region have been added. Thus, the efficiency is set to (n/(n+4))2 considering that all the particle beams outside the radiation exposure region 3 are wasted. Therefore, it can be understood that n only needs to set equal to 5 or more to adjust the efficiency of using particle beams to 30%. In the foregoing, the case where the radiation exposure region 3 was a circle was described. Generally, however, the radiation exposure region 3 is not a circular radiation field. Thus, in the method of the related art, the efficiency of using particle beams was lower than the theoretically value 30%. According to the second embodiment, since particle beams are applied only to the necessary belts outside the actual radiation exposure region 3, it is possible to increase the efficiency of using particle beams even if the radiation exposure region 3 is divided into the number of belts lower than five. Moreover, in the foregoing description, the radiation exposure region 3 and the peripheral region thereof were divided in parallel with the X axis with respect to the treatment center as shown in FIG. 5. However, a similar advantage can be obtained even if the division is made in parallel with a given straight line direction. In addition, the division was made in parallel with the X axis as shown in FIG. 5. However, a similar advantage can be obtained even if the division is made in parallel with a given curve. In addition, the analysis was carried out assuming that the beam shape (standard deviation σxy) was isotropic as shown in FIG. 5. However, even if the beam shape is not isotropic, simulation for optimizing a stepsize can be carried out according to the beam shape, providing a similar advantage. Further, the case where the radiation exposure region 3 was circle was described. However, a similar advantage can be obtained by carrying out simulation for applying particle beams to the region expanded by two belts outside the outermost belt in contact with the radiation exposure region 3, even if the radiation exposure region 3 has a given shape. Furthermore, the case where the flatness of the radiation exposure region 3 was ±1% was described. However, a similar advantage can be obtained even if flatness is optionally set. As apparent from FIG. 6, generally, the interval of the belts is smaller as the flatness of the radiation exposure region 3 is improved. As flatness is degraded, the interval of the belts is larger. If the flatness of the radiation exposure region 3 is improved, the dose of particle beams applied to the belt located outside from the end of the radiation exposure region 3 is increased. However, the dose of particle beams applied to the belt located inside the radiation exposure region 3 is also increased. On the other hand, if the flatness is degraded, the dose of particle beams applied to the belt located outside from the end of the radiation exposure region 3 is reduced. However, the dose of particle beams applied to the belt located inside the radiation exposure region 3 is also reduced. As apparent from the foregoing relation, the efficiency of using particle beams is not so dependent on the simulated flatness, if the range of flatness used for the radiation exposure treatment is about 0.5–5%, and standard deviation σxy of the beam shape is about 1.8–2.3. In addition, in the example shown in FIG. 5, there was no mention of the order of irradiating the belts with particle beams. However, since the influence of the irradiating order of the belts is not so important for the formation of the radiation exposure region 3, a similar advantage can be obtained irrespective of the order of the belts facilitating the control of the wobbler device or the order facilitating control data creation. In addition, in the example shown in FIG. 5, there was no mention of the direction of irradiating the belts with particle beams. However, since the influence of the irradiating direction of the belts is not so important for the formation of the radiation exposure region 3, a similar advantage can be obtained irrespective of the irradiating direction of the belts facilitating the control of the wobbler device 8 or the irradiating direction facilitating control data creation. Moreover, if the contribution of the dose of particle beams applied to each belt is equal among the belts, then a similar advantage can be obtained even if scanning is carried out not just once but by a plurality of times. In addition, in the example shown in FIG. 5, the division was made into the belts of one direction. However, a similar advantage can be obtained even if a plurality of belts divided in different directions are superposed. If the plurality of belts are superposed, the irradiating condition of each of the belts is relaxed. Examples may include a method of superposing the belts divided in the X and Y axes, a method of superposeing the belts divided in the X axis and along a straight line having a given angle in the X axis, and so on. If the dose of particle beams applied on the center line of each belt is equal, then a similar advantage is obtained even when movement is made so as to set constant the product of the speed and the number of particles of particles on each belt. In addition, the example of division into the belts coincident with the end of the circular radiation exposure region 3 was described. However, even if the belts are not coincident with the end of the radiation exposure region 3, a similar advantage can be obtained by simulating the dose of particle beams applied to the belt located outside the radiation exposure region 3 and including the belt contributing to the radiation exposure region 3. The example of irradiating the regularly disposed belts with the equal dose of particle beams having similar beam shapes was described. However, though calculation becomes a complex equation, even when the belts having irregular widths are irradiated with particle beams having different beam shapes, a similar advantage can be obtained if there is a solution to a reverse-problem of flattening the inside of the radiation exposure region 3. Further, assuming that the position to be irradiated with particle beams is changed timewise, it is possible to obtain the dose of particle beams applied to each belt by the foregoing method for the actual position of each belt to be irradiated with particle beams, if there is reproducibility or regularity in such a change. On the other hand, if there is no reproducibility or regularity in the change, then it is possible to obtain the flattening condition of the radiation exposure region 3 by including the width of the change in the beam shape of particle beams to be applied. As described above, according to the second embodiment, the radiation exposure region 3 and the peripheral region thereof to be irradiated with particle beams are divided in belt forms, the treatment simulation for applying particle beams is carried out according to each divided belt. During the execution of the treatment simulation, the radiation treatment condition is obtained for causing the flatness of the radiation exposure region 3 to be in a desired range, and the dose of particle beams applied to the belt of the peripheral region to be minimized, the radiation treatment plan reflecting the radiation treatment condition is made and, then, based on this treatment plan, particle beams are applied to the radiation exposure region 3 and the peripheral region thereof to be irradiated. Thus, since the unit radiation exposure regions obtained by dividing the radiation exposure region 3 and the peripheral region thereof into the plurality of regions are used, the radiation exposure region to be provided with desired flatness can be accurately decided, making it possible to increase the efficiency of using particle beams more compared with that in the case of the related art. Moreover, since particle beams are applied to the minimum region to be provided with desired flatness, it is possible to suppress the generation of superfluous particle beams. The radiation treatment system of the third embodiment is basically similar in configuration to that of the first embodiment. Thus, portions different from those of the first embodiment will be described. First, the treatment planning device 4 (simulation means and irradiation planning means) of the third embodiment forms a radiation exposure region 3 according to the state of the diseased part of the patient 2 as in the case of the first embodiment, divides the radiation exposure region 3 and the peripheral region thereof in concentric circular forms, and then sets a control parameter (radiation treatment condition) of each device for applying particle beams according to the center line of the interval of the concentric circles (unit radiation exposure region) in the accelerator controller 5 and the irradiation controller 6. The accelerator controller 5 and the irradiation controller 6 execute treatment simulation for applying particle beams according to the center line of the interval of the concentric circles divided according to the above control parameter. In the treatment simulation, the treatment planning device 4 obtains a control parameter of each device for causing the flatness of the radiation exposure region 3 to be in a desired range, and the dose of particle beams applied to peripheral region to be minimized, and then makes a radiation treatment plan reflecting this control parameter. Similarly to the case of the first embodiment, the treatment region 1 of the third embodiment is assume to be a ball having a radius r, and the center thereof is located at the cubic center of the patient 2 assumed to be a cube having one side 4r. FIG. 8 shows the example of dividing a radiation exposure region in the radiation treatment system of the third embodiment. As shown, the radius of an innermost circle is increased by a/2; and the radius of an outermost circle by a. Places to be irradiated with particle beams are on circles indicated by broken lines located on the center of a concentric circle and between concentric circles, and each radius is an integral multiple of a. Now, before the explanation of the operation of the radiation treatment system of the third embodiment, a condition for the flattening of the radiation exposure region 3 (referred to as the flattening of the radiation field, hereinafter) will be described. According to the third embodiment, assuming that particle beams are moved in such a way as to irradiate an exposure region between the concentric circles with the equal dose of particle beams per unit time, the treatment planning device 4 evaluates flatness inside the radiation exposure region 3 by using the equation described above with reference to the first embodiment. However, since the equation (3) assumes the movement of particle beams on a circle, an equation (5) below is used instead:Xi2+Yj2=(ka)2k=0, 1, 2, 3 . . . (5) As in the case of the first embodiment, if it is assumed that the beam shape of particle beams is subjected to two-dimensional Gaussian distribution, then, by obtaining the value of the equation (2) based on the condition of the equation (5), it is possible to calculate a dose distribution when the same dose is applied at the intervals of respective concentric circles. For the condition of flatness, an advantage almost similar to that of the first embodiment can be expected. Accordingly, a relational equation can be obtained between the beam shape and the radius of the concentric circle according to flatness. As in the case of each of the first and second embodiments, if a relation between the interval of the concentric circles standardized by the beam shape (standard deviation σxy) and flatness is assumed, then particle beams must be applied at the interval of at least two concentric circles outside the radiation exposure region 3. This is attributed to the fact that only by the dose from the end of the radiation exposure region 3, provided from particle beams moved on the outermost concentric circle inside the radiation exposure region 3, it is impossible to obtain desired flatness in the radiation exposure region 3. In other words, desired flatness can be obtained by contribution (1.9σxy) of particle beams on the concentric circle rotated in the nearest outside from the end of the radiation exposure region 3, and contribution (3.8σxy) of particle beams on the concentric circle rotated in the outside farther separated away from the end of the radiation exposure region 3. On the other hand, contribution (5.7σxy) of particle beams rotated on the concentric circle located in the outside yet farther separated away from the end of the radiation exposure region 3 becomes a size to be ignored. In this case, the application of the equal dose of particle beams per unit time means a condition for causing the moving speeds of particle beams on the concentric circles to be equal to one another. Accordingly, the rotational time of particle beams applied to the concentric circle of the outside is lengthened, making it possible to reduce a frequency for rotating particle beams in inverse proportion to the rotational radius. As for the condition for obtaining the desired flatness of the radiation field, and minimizing the number of concentric circles in the peripheral region of the radiation exposure region 3 (minimizing the dose of particle beams applied to the peripheral region), a specific control parameter therefor is calculated by the treatment planning device 4 during the treatment simulation. Next, the operation of the system as configured above will now be described as below. First, before the execution of particle-beam treatment, the diseased part (equivalent to the treatment region 1) of the patient 2 is photographed by the X-ray CT, not shown, and the obtained image data of the diseased part is output to the treatment planning device 4. Based on the state of the diseased part analyzed from the input image data of the diseased part, the treatment planning device 4 decides a radiation exposure region 3 by adding a region as a margin to the treatment region 1. In this case, it is assumed that in the treatment simulation for forming the region to be irradiated with particle beams, carried out at the treatment planning device 4 according to the third embodiment, as shown in FIG. 8, the radiation exposure region 3 and the peripheral region thereof are divided in concentric circular forms, and the equal dose of particle beams is applied on the center line at the interval (unit radiation exposure region) of the concentric circles. The treatment planning device 4 calculates a control parameter for applying particle beams on the center line at the interval of the concentric circles, and then outputs the control parameter to the accelerator controller 5 and the irradiation controller 6. Based on the control parameter input from the treatment plan device 4, the irradiation controller 6 sets leaf-control parameters for the collimator device 13 to be circumscribed on the radiation exposure region 3 assumed to be a circle having a radius r, and controls the movements of the wobbler device 8, the scatterer device 9, the dose monitoring device 10, the ridge filter device 11, and the range shifter device 12. Associatively, the accelerator controller 5 controls the accelerator 7, generates particle beams having a beam shape decided by the interval of the concentric circles shown in FIG. 8, and then executes treatment simulation (simulation step). In this case, to apply particle beams in a concentric circular form, the energizing current of the wobbler device 8 is controlled, and a rotational magnetic field for providing a prescribed angle of deflection to particle beams is generated. In the treatment simulation carried out in the foregoing manner, the treatment planning device 4 determines which of the concentric circles located in the outside from the end of the radiation exposure region 3 is irradiated with particle beams, in order to obtain desired flatness inside the radiation exposure region 3. Moreover, in the foregoing case, no particle beams need to be applied to a portion located away by three concentric circles or more from the end of the radiation exposure region 3. Thus, a coordinate of the concentric circle to be irradiated with particle beams can be obtained by the treatment simulation. Then, the treatment planning device 4 calculates a frequency of particle beams when one rotation is made on each concentric circle, and the dose of particle beams to be applied by one rotation. Subsequently, by using the dose monitoring device 10, the treatment planning device 4 determines whether the dose distribution of the particle beams applied according to the control parameter is appropriate or not. If the appropriate dose is determined, then, the treatment planning device 4 makes a radiation treatment plan reflecting the control parameter of the concentric circle to be irradiated with particle beams (radiation treatment planning step). The control parameter of the above radiation treatment plan is output from the treatment planning device 4 to the accelerator controller 5 and the irradiation controller 6. The accelerator controller 5 and the irradiation controller 6 set the same control parameter as that of the treatment simulation input from the treatment planning device 4 in the accelerator 7, the wobbler device 8, the scatterer device 9, the dose monitoring device 10, the ridge filter device 11, and the range shifter device 12. In addition, leaf control data corresponding to the radiation exposure region 3 is calculated, and set in the collimator device 13. Subsequently, the patient 2 is laid on the treatment couch 4 and fixed, and the radiation exposure region 3 is aligned with the radiation exposure position. When the radiation exposure region 3 is irradiated with particle beams, the irradiation controller 6 controls the wobbler device 8 to set particle beams in the position corresponding to the concentric circle to be irradiated with particle beams, sets a scatterer for providing a prescribed beam shape in the scatterer device 9, and then sets the dose of particle beams to be applied in the dose monitoring device 10 according to the radiation treatment plan. After the setting of the control parameter in the foregoing manner, the application of particle beams at the interval of the concentric circles is started. When the dose of particle beams monitored by the dose monitoring device 10 reaches a prescribed value (particle beams make one rotation), the irradiation controller 6 stops the radiation beam, and performs control to move the particle beams to a next concentric circle, and then the application of the particle beams is continued (radiation exposure step). After the foregoing operation has been carried out for all the intervals of the concentric circles, one radiation exposure treatment is completed. According to the third embodiment, for the flattening of the radiation field, if no particle beams are applied to the peripheral region within the interval of three concentric circles or less from the end (boundary) of the radiation exposure region 3, flatness inside the radiation exposure region 3 cannot be secured. Therefore, a part of particle beams applied to the interval of two concentric circles from the end (boundary) of the radiation exposure region 3 is wasted. In other words, assuming that the radiation exposure region 3 is a circle having a radius r, and its diameter 2r is equivalent to n concentric circles, it is possible to approximate the number of the intervals of concentric circles necessary for the flattening of the radiation field by the area of a circle having a diameter (n+4), formed by adding the interval of two concentric circles from the end of the radiation exposure region 3. The efficiency of using particle beams is given by the ratio of the number of concentric circles in the radiation exposure region 3, and the number of concentric circles in the region obtained by adding two concentric circles of the peripheral region to the radiation exposure region 3. Thus, the efficiency is given by (n/(n+4))2, considering that all the particle beams outside the radiation exposure region 3 are wasted. Therefore, it can be understood that to adjust the efficiency of using particle beams to 30%, n must be increased to 5 or more. Moreover, in the foregoing, the case where the radiation exposure region 3 was a circle was described. Generally, however, the radiation exposure region 3 is not a circular radiation field. Therefore, the efficiency of using particle beams was lower than the theoretical value 30% in the method of the related art. According to the third embodiment, by making division in a flat concentric elliptic form according to the actual radiation exposure region 3, it is possible to increase the actual efficiency of using particle beams even if the radiation exposure region 3 is divided into five concentric ellipse or higher. Moreover, in the foregoing, the case where the radiation exposure region 3 and the peripheral region thereof were divided in the odd number of concentric circles so as to move the particle beams on the boundary of the radiation exposure region 3 as shown in FIG. 8. However, a similar advantage can be obtained by setting the even number of divided regions so as to prevent the movement of the particle beams on the boundary of the radiation exposure region 3, passing the middle part between radiation beams on the boundary. In addition, in the foregoing, the example of division into the concentric circles was described. However, a similar advantage can be obtained even if division is made into given concentric ellipses. In addition, the analysis was carried out assuming that the beam shape (standard deviation σxy) was isotropic as shown in FIG. 8. However, even if the beam shape is not isotropic, simulation for optimizing a stepsize according to the beam shape can be performed, proving a similar advantage. The case where the radiation exposure region 3 was a circle was described. However, a similar advantage can be obtained by executing simulation for applying particle beams to the region expanded by two concentric circles outside the outermost concentric circle in contact with the radiation exposure region 3, even for the radiation exposure region 3 having a given shape. In addition, the case where the flatness of the radiation exposure region 3 was ±1% was described. However, a similar advantage can be obtained even for given flatness. Generally, the interval of the concentric circles is smaller as the flatness of the radiation exposure region 3 is improved. The interval of the concentric circles is larger as the flatness is degradded. By improving the flatness of the radiation exposure region 3, the dose of particle beams applied to the concentric circle located outside from the end of the radiation exposure region 3 is increased. However, the dose of particle beams applied to the concentric circle located inside the radiation exposure region 3 is also increased. On the other hand, if the flatness is degraded, the dose of particle beams applied to the concentric circle located outside from the end of the radiation exposure region 3 is reduced. However the dose of particle beams applied to the concentric circle located inside the radiation exposure region 3 is also reduced. Apparently from the foregoing relation, the efficiency of using particle beams is not so dependent on the simulated flatness if the range of flatness used for the radiation exposure treatment is about 0.5–5%, and the standard deviation σxy of the beam shape is about 1.7–2.1. Further, in the example shown in FIG. 8, there was no mention of the order of irradiating the interval of the concentric circles with particle beams. However, since the influence of the irradiating order of the concentric circles is not so important for the formation of the radiation exposure region 3, a similar advantage can be obtained irrespective of the irradiating order of the concentric circles facilitating the control of the wobbler device 8 or the irradiating order facilitating control data creation. In the foregoing, the center portion was not wobbled. However, even when the center portion is wobbled as in the case of the method of the related art, a similar advantage can be obtained by carrying out similar simulation. Moreover, if the contribution of the dose of particle beams is similar among the concentric circles, then a similar advantage can be obtained even when rotation is made not just once but by a plurality of times. If the dose of particle beams is equal among the concentric circles, then a similar advantage can be obtained even when movement is made so as to set constant the product of the speed of particle beams and the number of particle beams. The example where the concentric circles were regularly disposed, and the same dose of particle beams having similar beam shapes were applied in each concentric circle was described. However, though calculation becomes a complex equation, even when particle beams having different beam sizes are applied to the irregularly disposed concentric circles, a similar advantage can be obtained if there is a solution to a reverse-problem for flattening the inside of the radiation exposure region 3. In addition, assuming that the position to be irradiated with particle beams is changed timewise, if there is reproducibility or regularity of such a change, then it is possible to calculate the dose of particle beams applied to the position by the foregoing method. On the other hand, if there is not reproducibility or regularity of the change, then it is possible to obtain the flattening condition of the radiation exposure region 3 by including the width of the change in the beam size of particle beams to be applied. As described above, according to the third embodiment, the radiation exposure region 3 and the peripheral region thereof to be irradiated with particle beams are divided in the concentric circular forms, the treatment simulation is executed according to the interval of the divided concentric circles. During the treatment simulation, the radiation treatment condition is obtained for causing the flatness of the radiation exposure region 3 to be in a desired range, and the dose of particle beams applied to the interval of the concentric circles of the peripheral region to be minimized, the radiation treatment plan reflecting this radiation treatment condition is made and, based on the radiation treatment plan, the radiation exposure region 3 and the peripheral region thereof are irradiated with particle beams. Thus, since the unit radiation exposure regions obtained by dividing the radiation exposure region 3 and the peripheral region into the plurality of regions, it is possible to accurately decide the radiation exposure region for providing desired flatness, and to increase the efficiency of using particle beams more compared with the case of the related art. Moreover, since particle beams are applied to the interval of the concentric circles of the minimum peripheral region for providing desired flatness, it is possible to suppress the generation of superfluous particle beams. It is necessary to supply much more current to the wobbler device 8 when particle beams are rotated outside. However, if rotation is made so as to set constant the dose of particle beams applied to each concentric circle, the frequency of the concentric circle outside is reduced in inverse proportion to the rotational radius, lowering a power supply load or the like on the wobbler device 8. Thus, it is possible to facilitate system configuration. As described heretofore, according to the present invention, the radiation exposure region and the peripheral region thereof are divided into a plurality of unit exposure regions, and the radiation treatment simulation is executed for applying particle beams according to the shape of each divided unit radiation exposure region. During the radiation treatment simulation, the radiation treatment condition is obtained for causing the flatness, i.e., the degree of uniformly irradiating the radiation exposure region with a proper dose of particle beams, to be in a desired range, and the dose of particle beams applied to the unit radiation exposure region of the peripheral region to be minimized, and then the radiation treatment plan reflecting this radiation treatment condition is made. Thus, it is possible to accurately decide the radiation exposure region for providing desired flatness, and to increase the efficiency of using particle beams more compared with the case of the relate art. Moreover, since particle beams are applied to the minimum peripheral region for providing desired flatness, it is possible to suppress the generation of superfluous particle beams. According to this invention, the radiation exposure region and the peripheral region thereof are divided into the unit radiation exposure regions of the grid form. Thus, it is possible to accurately decide the radiation exposure region for providing desired flatness. According to the invention, the radiation exposure region and the peripheral region thereof are divided into the belt-like unit exposure regions. Thus, an advantage similar to the above can be obtained by the foregoing method. According to the invention, since the radiation exposure region and the peripheral region thereof are divided into the concentric circular unit exposure regions, an advantage similar to the above can be obtained by the foregoing method. Moreover, since rotation is made so as to set constant the dose of particle beams applied to each concentric circle, the frequency of the concentric circle outside is reduced in inverse proportion to the rotational radius. Therefore, it is possible to reduce a power supply load or the like on the wobbler device. Furthermore, according to the invention, when the unit exposure region is located in the boundary of the radiation exposure region, determination is made as to the degree of contribution made by the dose of particle beams applied to the unit exposure region located in the boundary to the radiation exposure region, according to the dose of particle beams applied to the unit exposure region located in the boundary of the peripheral region. Thus, it is possible to suppress the generation of superfluous particle beams. |
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053944496 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The impact limiters formed in accordance with the present invention are used in the transport of large casks that carry canisters which contain a plurality of spent nuclear fuel assemblies. The spent nuclear fuel assemblies are elongate rods that are carried within the canisters by a basket. The canister serves as a sealed enclosure around the basket. The canister is generally in the shape of a right cylinder. In addition to its use as a container to transport spent nuclear fuel over public thoroughfares, such canisters and baskets are employed to store and transfer spent nuclear fuel on-site, e.g., at a nuclear power plant. One such canister and basket combination is described in an application entitled Containers for Transportation and Storage of Spent Nuclear Fuel, filed on Oct. 8, 1993 and assigned U.S. application Ser. No. 08/131,971, naming Robert A. Lehnert, Robert D. Quinn, Steven E. Sislcy, and Brandon D. Thomas as inventors. The subject matter of the above-identified application is expressly incorporated herein by reference. When the canisters containing the spent nuclear fuel are ready for transportation off-site over public thoroughfares, the canisters are removed from their temporary storage facility and placed within a transportation cask, which is generally a sealed steel vessel in the shape of a right cylinder. The canister is received within the transportation cask, which is then sealed. One such transportation cask for transporting the canister is described in an application entitled Transportation and Storage Cask for Spent Nuclear Fuel, filed on Oct. 8, 1993, and assigned U.S. application Ser. No. 08/131,973, naming Kyle B. Jones, Robert E. Lehnert, Ian D. McInnes, Robert D. Quinn, Steven E. Sislcy, and Charles J. Temus as inventors. The subject matter of the above-identified application is expressly incorporated herein by reference. Referring to FIG. 1 of the subject application, a transportation cask (phantom lines in FIG. 2) carrying the canister/basket combination is enclosed by a skid 20, and front impact limiter 22 and rear impact limiter 23 formed in accordance with the present invention rests horizontally on conventional trailer 26. In FIG. 1, the transportation cask is not visible, as it is completely encased by skid 20 and impact limiters 22 and 23. Skid 20 is further enclosed by a curtain of expanded metal 24, which further obscures skid 20 and the transportation cask. The curtain of expanded metal 24 is provided around skid 20 in order to shield skid 20 and the transportation cask from sunlight. In FIG. 1, the longitudinal axis of the transportation cask is parallel to the length of trailer 26. Front impact limiter 22 is positioned on the forward end of the generally cylindrical transportation cask, and rear impact limiter 23 is positioned on the opposing rearward end. Skid 20 supports the transportation cask along its length between impact limiters 22 and 23. In accordance with the present invention, impact limiters 22 and 23 include peripheral outer surfaces that are non-circular, e.g., multisided. In the illustrated embodiment, impact limiters 22 and 23 are generally octagonal elements that include an annular, generally octagonal body 28 and a tapered cap 30 which has a periphery in a cross section perpendicular to the longitudinal axis of the transportation cask that approximates an octagon. The outer surface of impact limiters 22 and 23 are essentially identical. The following description regarding impact limiter 22 is equally applicable to impact limiter 23. The same reference numerals for similar elements are used for impact limiters 22 and 23. In the illustrated embodiment, the exterior of annular body 28 includes four primary sides 32, each separated by a smaller secondary side 34. The exterior of tapered cap 30 is a truncated conical member with a plurality of sides. In the following description of the illustrated embodiment, representative dimensions are provided; however, it should be understood that these relative dimensions may be varied, depending on the particular size of the transportation cask and other design factors, such as impact resistance, space envelope available, and construction materials. In addition, the following description includes exemplary materials of construction, which also can be varied due to a number of concerns, such as those listed above. Referring additionally to FIGS. 2A, 4A, and 5, primary side 32 and secondary side 34 of impact limiter 22 each have a width extending in the direction parallel to the longitudinal axis of the transportation cask that is equal. The length of primary side 32 in a direction transverse to the longitudinal axis of the transportation cask is greater than the length of secondary side 34 in a direction transverse to the longitudinal axis of the transportation cask. In the illustrated embodiment, primary side 32 is approximately four times longer than the length of secondary side 34. In the illustrated embodiment, the length of primary side 32 is approximately 95 inches and the length of secondary side 34 is approximately 23 inches. The width of primary side 32 and secondary side 34 in a direction parallel to the longitudinal axis of the transportation cask is approximately 35.5 inches. Referring additionally to FIG. 3, tapered cap 30 of impact limiter 22, like annular body 28, includes an outer shell defined by four primary sides 36 each separated by a secondary side 38. Tapered cap 30 also includes a substantially square end plate 40 opposite annular body 28. Tapered cap 30 extends in a direction parallel to the longitudinal axis of the transportation cask from an end of annular body 28 opposite the transportation cask. Square end plate 40 is spaced apart from annular body 28 along the longitudinal axis of the transportation cask. End plate 40 has sides whose lengths are less than the length of primary sides 36. In the illustrated embodiment, the sides of end plate 40 are about 70 inches long and end plate 40 is spaced apart from annular body 28 about 49 inches. The parallel sides of trapezoidal-shaped primary sides 36 are defined by the edge of primary sides 36 opposite the transportation cask and one side of square end plate 40. The longer of the two parallel sides corresponds to the edge of primary side 36. The nonparallel ends of trapezoidal-shaped primary sides 36 are equal in length. Secondary sides 38 are triangular and include a base defined by the edge of secondary side 34 opposite the transportation cask and by one nonparallel end from the two primary sides 36 that are on either side of secondary side 38. Accordingly, the combination of primary side 36, secondary side 38 and end plate 40 provides a tapered cap 30 in the shape of a truncated conical element having an outer periphery in a plane parallel to end plate 40 that approximates an octagon. Referring additionally to FIGS. 2B and 4B, rear impact limiter 23 is illustrated and as noted above includes an outer surface that is substantially identical to the outer surface described above with respect to impact limiter 22. Referring to FIG. 4A, annular body 28 of impact limiter 22 includes an outer shell 42 comprising primary sides 32 and secondary sides 34, described above, and a round inner shell 44 that is concentrically located relative to outer shell 32. In the embodiment illustrated in FIG. 4A, inner shell 44 is circular and has a diameter of approximately 84 inches and a height of about 32.5 inches. The end of inner shell 44 opposite tapered cap 30 includes twelve substantially equally-spaced notches (45 in FIG. 5) that, in the illustrated embodiment, are approximately three inches wide and three inches deep. The spacing between notches 45 is such that they are aligned with notches 49, 49a, 49b, and 49c and bolt conduits 56 described below for securing impact limiter 22 to a transportation cask. Inner shell 44 and outer shell 42 define an annular space therebetween. It should be understood that the respective sides and various elements of annular body 28 and tapered cap 30 can be formed by connecting individual plates of metal, such as type 304 stainless steel having a thickness of about 1/4 inch by conventional means such as welding. The end of annular body 28 opposite end cap 40 is closed by cover plate 48, which is an annular ring extending between outer shell 42 and inner shell 44. Cover plate 48 has an outer periphery that substantially matches the outer periphery of annular body 28. The inner periphery of cover plate 48 includes a bore having a diameter which in the illustrated embodiment is approximately 84 inches. This bore includes substantially square notches 49, 49a, 49b, and 49c for allowing bolt conduits 56 (described below in more detail) to pass through cover plate 48. In the illustrated embodiment, notches 49 are spaced apart from each other about 30 degrees. Notches 49a are spaced from the adjacent notches 49 about 30 degrees and notches 49b by about 35 degrees. Notches 49b are spaced from notches 49c by about 25 degrees. Notches 49c are spaced from adjacent notches 49 about 30 degrees. The variable spacing between notches 49, 49a, 49b, and 49c in impact limiter 22 is not required and other spacings such as equal spacing are within the scope of the present invention. Cover plate 48 around its outer periphery includes a 90.degree. elbow of metal to which outer shell 42 of annular body 28 can be secured by conventional means such as welding. Attached to the end of inner shell 44, opposite cover plate 48, is a circular plate 50. Circular plate 50 is centered along the axis of inner shell 44. In the illustrated embodiment plate 50 has a diameter of about 91 inches. Extending at a 90.degree. angle from the periphery of plate 50 towards end plate 40 is a lip of metal that forms shear stop (52 in FIG. 5). Shear stop 52 provides a lip that acts to oppose the shearing of the foam impact absorbing material along the plane where the honeycomb impact absorbing material and the foam impact absorbing material meet. In the illustrated embodiment, plate 50 is a type 304 stainless steel having a thickness of about 0.5 inch. Shear stop 52 is formed from the same 0.5 inch thick 304-type stainless steel and has a length of approximately 3.5 inches. Circular plate 50 adjacent its periphery includes twelve substantially equally-spaced bores for allowing bolts (58 in FIG. 5) to pass therethrough. The spacing between these bores is such that they can be aligned with notches 49, 49a, 49b, and 49c. Though not illustrated, plate 50 on its exposed surface includes a means for isolating the transportation cask from plate 50. One way of accomplishing this isolation is through the attachment of a concentric spacer ring approximately 1/4 inch high and 1/4 inch wide to plate 50. It is desirable to space the transportation cask from plate 50 in order to decrease the risk that heat from transportation cask will cause plate 50 to become hot enough so as to damage the underlying foam, which is described below in more detail. Conversely, spacing the transportation cask from plate 50 provides an air gap that protects the seals on the transportation cask from damage caused by overheating should the exterior of the impact limiter be exposed to high temperatures, for example, during a fire incident. Secured to the outer surface of inner shell 44 within the annular space between inner shell 44 and outer shell 42 are twelve bolt conduits 56 for receiving bolts 58. Bolt conduits 56 are essentially tubular elements which in the illustrated embodiment have a length of approximately 32.75 inches. Bolt conduits 56 include a receiving end for receiving bolt 58 which is secured to the exposed surface of plate 50 aligned with one of the twelve bores in plate 50 described above. Bolt conduits 56 extend to cover plate 48 and are attached to cover plate 48 within notches 49, 49a, 49b, and 49c and aligned with notches 45. Bolt conduits 56 are essentially three-sided elongate members with the three sides forming three quarters of a square tube. The fourth side of the square tube which defines bolt conduits 56 is provided by inner shell 44. As briefly noted above, bolt conduits 56 provide a passageway for a plurality of longnecked bolts 58 that serve to secure impact limiter 22 to the transportation cask as described below in more detail. Notches 45 and bolt conduits 56 are sized to receive lugs identified by reference numeral 53 on the transportation cask in FIGS. 2 and 5. The cooperation between lugs and the pocket formed by bolt conduits 56 and notches 45 serves to prevent rotation of the impact limiter relative to the cask. This serves to reduce shear stress on the bolts which otherwise would be the sole means of preventing rotation in the absence of the lugs and notch combination. Situated within each bolt conduit 56 in the opening adjacent cover plate 48 is a bolt guide 60 that serves to align the threaded ends of a bolt 58 with a corresponding threaded female member in lugs 53 on the transportation cask. Bolt guides 60 are square plates having a centrally located bore passing therethrough. Bolt guide 60 is dimensioned to fit within and be secured to bolt conduit 56. Bolt guides 60 are secured within bolt conduits 56 by conventional means, such as welding. As noted above, impact limiter 23 is substantially identical to impact limiter 22. Referring to FIGS. 2B AND 4B inner shell 44 of impact limiter 23 at locations offset 180.degree. from each other includes a rectangular slot 104 extending from the end of inner shell 44 adjacent cover plate 48 to a position approximately half way into the bore that receives the transportation cask. Slot 104 opposite cover plate 48 includes an arcuate end. Slot 104 disrupts the generally circular shape of inner shell 44 for impact limiter 23 and provides space for receiving structural elements such as support trunions that are located on the transportation cask. Impact limiter 22 does not include such a slot as the end upon which impact limiter 22 is positioned does not include trunions at a location where they would be received within impact limiter 22. In the illustrated embodiment, slot 104 is approximately 19 inches wide and 10 inches long. Referring primarily to FIGS. 3 and 5, tapered cap 30 in the illustrated embodiment includes twelve substantially equally-spaced conduits 62 for receiving bolts 58. The spacing between conduits 62 corresponds to the spacing of notches 49, 49a, 49b, and 49c described above. Conduits 62 are axially aligned with bolt conduits 56 and in the illustrated embodiment are thin-walled tubes of 300 series stainless steel having a diameter of approximately 2" and a length sufficient to allow them to extend from the outer shell of tapered cap 30 to the unexposed side of plate 50. Conduits 62 are centered on a circle having a diameter that coincides with the diameter of the circle that bolt conduits 56 are centered within annular body 28. Accordingly, as illustrated in FIG. 3, four bolt conduits 62 extend from end plate 40 to plate 50 and eight bolt conduits 62 extend from respective primary sides 36 to plate 50. As described above, plate 50 at locations aligned with bolt conduit 62 includes 12 apertures for allowing the narrow neck of bolts 58 to pass therethrough while preventing the heads of bolts 58 from passing therethrough. Accordingly, impact limiter 22 is secured to the transportation cask using bolts 58 that are not directly exposed to impacts which could exert high shear forces on the bolts. Outer shell of tapered cap 30 includes a number of other ports. End plate 40, in addition to bolt conduits 62, includes four safety valves 66 that are designed to rupture should pressure within impact limiter 22 exceed a predetermined level. Examples of situations where pressure within impact limiter 22 may exceed predetermined levels include heating of the foam within impact limiter 22 caused by a rising external temperature. The build-up of pressure within impact limiter 22 can result in a blow-up of the impact limiter unless the gas is vented. Safety valves 66 in the illustrated embodiment are pipe fittings which have been secured within bores passing through outer shell 30 by conventional means, such as welding. The pipe fittings are plugged with a conventional plastic pipe plugs that soften with rising temperature. Softening of the pipe plugs combined with pressure build-up within impact limiters 22 or 23 causes the plugs to rupture and relieve pressure within the limiters. Though not required, additional safety valves 64 can be provided in primary sides 36. Located at the four corners of end plate 40 are vent holes 68 which, as described below in more detail, are present in order to relieve pressure buildup during the pouring of the foam and its actual foaming. Centrally located within end plate 40 is an access port 70 through which foam is poured as described below. Referring primarily to FIG. 5, in order for impact limiters 22 and 23 to absorb energy during impact and protect the transportation cask, an impact absorbing material 72 is contained within annular body 28 and a different and distinct impact limiting material 74 is located within tapered cap 30. The use of two different and distinct types of impact absorbing material provides a means to take full advantage of the crush properties of more than one material. By using the different materials, optimum properties of each can be used to provide advantageous designs. For example, in the illustrated embodiment, a honeycomb material with cross-laminated corrugations is used as impact absorbing material 72 and a foam material is used as impact absorbing material 74. In FIG. 5, the honeycomb impact absorbing material 72 is depicted schematically without illustrating the cross-laminated corrugations. Also, FIG. 5 depicts a cross-section of impact limiter 22, it should be understood that the same cross-section of impact limiter 23 would include slot 104. The use of a honeycomb with cross-laminated corrugations in the annular body around the transportation cask where elevated temperatures may be encountered is advantageous because the honeycomb can be machined from materials that exhibit minimal temperature dependencies and thus will give more uniform results over a wider range of temperatures. In addition, honeycomb with cross-laminated corrugations exhibit crush properties that are desirable in locations where the space available for crushable material, is limited. Referring to FIGS. 5, 6 and 7 in the illustrated embodiment, the preferred honeycomb material is a multidirectional aluminum honeycomb which has an essentially uniform crush strength in any orientation within a plane and a lesser strength in a direction perpendicular to that plane. This type of honeycomb material differs from conventional honeycomb which is very directional in a given plane. Such honeycomb materials have corrugations that are cross-laminated in alternating layers. Exemplary types of such aluminum honeycomb material are available from Alcore, Inc., Belcamp, Md., under the name Trussgrid Honeycomb. Another type of aluminum honeycomb material is available from Hexcel Corporation, Dublin, Calif., under the name Cross-Core. An exemplary honeycomb material has a nominal static crush strength of 1,650 PSI, a minimum strength of not less than about 1,400 PSI when tested at 200.degree. F. and 67 feet per second dynamic crush rate (FPS), and a maximum strength of not more than about 1,950 PSI when tested at -20.degree. F. and 67 FPS. In the illustrated embodiment, the honeycomb material for impact limiter 23 is provided in the form of four distinct layers 76, 78, 80 and 82 laminated in an axial direction relative to the cask. Layers 76, 78, 80, and 82 are sized to fit between inner shell 44 and outer shell 42 of annular body 28. Layers 78, 80 and 82 include a rectangular cutout 106 on their inner periphery to account for slot 104. Though not illustrated, layers 76, 78, 80 and 82 for impact limiter 22 are identical to the same layers used in impact limiter 23 with the exception that rectangular cutout 106 to accommodate for slot 104 is unnecessary. In the illustrated embodiment, individual layers 76, 78, 80 and 82 of honeycomb material are each approximately 8.8 inches thick. Layers 76 and 80 are formed from four blocks 84 having a generally rectangular shape. Each of the four blocks 84 that form layers 76 and 80 are substantially identical with the exception that two blocks 84 of layer 80 include a cutout portion as described above. Each block 84 is made up of smaller sections of the aluminum honeycomb material. The joints between adjacent blocks 84 in layers 76 and 80 are formed by aluminum sheets (not shown) positioned in the joint with an epoxy resin providing the adhesion. In the illustrated embodiment, the aluminum sheet material is 28 gauge. Layers 78 and 82 are each composed of four generally triangular blocks 86 which are also joined to each other using aluminum divider sheets and epoxy resin. Block 86 of layer 82 includes a cutout portion as described above. Sheets of aluminum used to form the joints are also used to seal the cell ends for each layer 76, 78, 80 and 82 by encasing each layer around its periphery. Adjacent layers 76, 78, 80 and 82 are adhered to each other using epoxy resin and sheets of aluminum. In the embodiment illustrated in FIG. 6, sheets 81 of aluminum are cut to the shape of blocks 84 and sheets of aluminum 83 are cut to the shape of blocks 86. Alternatively, a single sheet of aluminum could be cut out to the shape of the individual layer in order to provide the bond line necessary to adhere adjacent layers together. Each layer 76, 78, 80 and 82 includes a centrally located bore 85 which is sized to allow the individual layers to encompass and surround inner shell 44. In addition, bore 85 includes notched cutouts 87 that mate with bolt conduits 56. Layer 82 of the honeycomb material differs slightly from layer 78 and includes a seat 89 adjacent bore 85 for mating with the shell of impact limiters 22 and 23. Layers 76 and 80 include joint lines between adjacent blocks 84 that are offset 45.degree. from the joint lines between adjacent blocks 86 of layers 78 and 82. By offsetting the joint lines, the risk of the cask wedging apart the individual blocks in all of the layers is reduced due to the discontinuity of the joint lines. The individual layers are adhered to each other using the same epoxy that connects the individual blocks. The formed block of honeycomb material has dimensions that allow it to fit within annular body 28; however, the block of honeycomb material is not secured or adhered to outer shell 42 or inner shell 44. In the illustrated embodiment, the block of honeycomb material on its outside periphery includes two rows of spaced-apart 0.75 inch deep grooves that make the contact surface between the honeycomb material and the inside surface of outer shell 46 approach approximately 45% of the overall outer surface area of the honeycomb block. By providing the grooves 88 around the honeycomb block, the resistance to crush force on the exterior of the annular body is reduced. In other words, grooves 88 tend to soften the impact limiter at least near the honeycomb/outer shell interface. Softening near this interface is often desirable so that "soft" impacts, for example, a one-foot drop, are readily absorbed without any substantial damage to the impact limiter or cask, while still providing an impact absorbing structure that will protect the cask when impacts of greater magnitude are encountered. The plane of strength of the honeycomb material should be oriented to be perpendicular to the longitudinal axis of the transportation cask. Because the aluminum honeycomb exhibits no strain hardening throughout its working crush range, when combined with a crush from the inside of the impact limiter, an essentially constant force-deflection crush is achieved. Depending on the particular orientation of the impact received by the impact limiter, the honeycomb material may absorb all or a portion of the impact. For example, a flat impact received on either primary side 36 or secondary side 30 will be absorbed primarily by the honeycomb. In contrast, a flat impact received on end plate 40 would be essentially absorbed by the foam material described below and the honeycomb material would not be active. As described above, tapered cap 30 of impact limiter 22 encloses an impact absorbing material 74 different from the impact absorbing material 72. In the illustrated embodiment, impact absorbing material 74 is a closed cell polyurethane foam of nominal 15 pounds per cubic foot density. A preferred foam has virtually isotropic properties in a predetermined plane. The closed cell polyurethane foam is fire resistant, has low water absorption properties, and has predictable crush properties. The foam is introduced into tapered cap 30 through access port 70. In order to avoid adhesion between the foam and the inside surface of tapered cap 30 and the impact absorbing material 72, a release compound such as wax is applied to these surfaces prior to introduction of the foam. By preventing adhesion between the foam material and the underlying honeycomb, bonding of the materials, which generally creates a significant uncertainty with respect to the effect of an impact on the bond and the crush properties of the bonded materials, is avoided. By creating a design where the bond is not required, there is no concern about the quality of the joint that is provided. After the foam is poured into tapered cap 30, it expands and fills in tapered cap 30. Once expansion of the foam is complete, vents 68 and access port 70 are sealed. Using foam in tapered cap 30 introduces a material that changes crush properties with a change in temperature. For example, if the impact limiter is dropped on end plate 40, as the temperature goes up, the foam softens, which allows the foam to absorb the impact over a longer distance or stroke which results in lower loadings on the cask due to the impact. While the foam tends to harden as temperatures fall, the structural features of the cask that the foam is designed to protect also harden with colder temperatures, and accordingly, while the loads at colder temperatures will be higher due to a decreased stroke, the structures which are to be protected also can withstand higher loadings due to the lower temperatures. Referring back to FIG. 1, the multisided design of an impact limiter formed in accordance with the present invention provides a plurality of surfaces for receiving an impact. Because such surfaces are generally planar, upon impact, a larger surface area is activated compared to an impact limiter which has a generally circular periphery. The "activated area" is the area of the impact absorbing material that will be crushed as the impact is absorbed. In other words, an impact limiter that has a circular periphery will receive an impact along some tangential line of contact and thus only a small area of the circular impact limiter will be initially activated. In order to compensate for this small area of activation, the stroke of the impact absorbing material necessary to absorb the force of the impact must be increased which results in a larger size for the impact limiter or the stiffness of the impact absorbing material must be increased which has the effect of increasing the potential load on the transportation cask. In contrast, when a plurality of substantially planar surfaces are provided, the area which is activated upon impact is likely to be larger. Activating larger areas upon impact is desirable so that more of the impact absorbing material can be utilized in order to absorb the force of the impact. By increasing the amount of area to be activated during a given impact, the length of the stroke necessary to absorb the impact can be reduced which reduces the overall size of the impact limiter. Also, the stiffness of the impact absorbing material can be reduced which reduces the loads on the casks as the impact is absorbed. Furthermore, by spreading the impact force out over a larger area on the exterior of the impact limiter, for certain orientations, the crush mechanism will occur from the inside out, rather than from the outside in. A crush mechanism from the inside out is desirable because of the large area of impact absorbing material that is activated. The tapered cap of the impact limiter allows for a longer stroke for drop orientations where the size of the impact limiter is not restricted. In addition, the tapered cap also keeps the end portion which extends beyond the cask from influencing side drops or corner drops significantly. By proper selection of the taper as indicated by the embodiment described above, the loads on the cask for oblique drops and the resulting slap-down of the initially nonimpacted end can be minimized. While the preferred embodiment of the invention has been illustrated and described, it will be appreciated that various changes can be made therein without departing from the spirit and scope of the invention. For example, different types of metals could be used for the various structural elements of the impact limiter and in the skid of the present invention. In addition, other types of honeycomb material or foam materials may be employed in accordance with the present invention. Finally, the dimensions described above are exemplary of the many different dimensions that could be employed depending on the particular size of the cask and the particular materials used to form the impact limiter in accordance with the present invention. |
summary | ||
description | The present application is a continuation of commonly owned and assigned U.S. patent application Ser. No. 10/716,193, now U.S. Pat. No. 7,117,119, entitled SYSTEM AND METHOD FOR CONTINUOUS ONLINE SAFETY AND RELIABILITY MONITORING filed Nov. 17, 2003, which claims priority under 35 U.S.C. §119(e) to U.S. Provisional Patent Application Ser. No. 60/491,999 filed Aug. 1, 2003, entitled: System and Method for Continuous Online Safety and Reliability Monitoring, both of which are incorporated herein by reference. This application relates to U.S. patent application Ser. No. 10/684,329, now U.S. Pat. No. 7,133,727, filed Oct. 10, 2003, of Van Dyk, et al.; entitled SYSTEM AND METHOD FOR CONTINUOUS ONLINE SAFETY AND RELIABILITY MONITORING. The present invention relates generally to control and monitoring systems, and more specifically to industrial safety and reliability control and monitoring systems. Modern industrial systems and processes tend to be technically complex, involve substantial energies and monetary interests, and have the potential to inflict serious harm to persons or property during an accident. Although absolute protection may not be possible to achieve, risk can be reduced to an acceptable level using various methods to increase an industrial system's safety and reliability and mitigate harm if an event, e.g., a failure, does occur. In the context of safety systems, one of these methods includes utilization of one or more safety instrumented systems (SIS). A safety instrumented system (SIS) is an instrumented system used to implement one or more safety instrumented functions (SIF), and is composed of sensors, logic solvers and final elements designed for the purposes of: taking an industrial process to a safe state when specified conditions are violated; permitting a process to move forward in a safe manner when specified conditions allow (permissive functions); and/or taking action to mitigate the consequences of an industrial hazard. As mentioned above, a safety instrumented function (SIF) is a function implemented by a SIS which is intended to achieve or maintain a safe state for a process with respect to a specific event, e.g., a hazardous event. Hardware to carry out the SIF typically includes a logic solver and a collection of sensors and actuators for detecting and reacting to events, respectively. To direct appropriate design and planned maintenance of a SIF, safety standards bodies have established a system that defines several Safety Integrity Levels (SIL) that are appropriate for a SIF depending upon the consequences of the SIF failing on demand. According to the International Electrotechnical Commission (IEC) standard 61508, safety integrity level (SIL) is a measure of the risk reduction provided by a SIF based on four discrete levels, each representing an order of magnitude of risk reduction. As shown in Table 1, each SIL level is associated with a designed average probability of failure on demand (PFD). For example, a SIL 1 means that the maximum probability of failure is 10% (i.e., the SIF is at least 90% available), and a SIL 4 means that the maximum probability of failure is 0.01% (i.e., the SIF is at least 99.99% available). TABLE 1DEMAND MODE OF OPERATIONSafety IntegrityTarget Average Probability ofLevel (SIL)Failure on DemandTarget Risk Reduction4≧10−5 to <10−4>10,000 to ≦100,0003≧10−4 to <10−3 >1000 to ≦10,0002≧10−3 to <10−2>100 to ≦10001≧10−2 to <10−1>10 to ≦100 For continuous or high demand mode of operation, the following Table 2 applies: TABLE 2CONTINUOUS MODE OF OPERATIONTarget Frequency ofDangerous Failures to perform theSafety Integritysafety instrumented function (perLevelhour)4≧10−9 to <10−83≧10−8 to <10−72≧10−7 to <10−61≧10−6 to <10−5 Consistent with existing, standardized methodology, during design of a safety instrumented system (SIS), safety integrity level (SIL) requirements are established for each SIF based upon the impact of the specific hazardous event that the SIF is intended to prevent. For example, a SIL level of 1 may be assigned to a hazardous event that imparts only minor property damage, whereas a SIL of 4 may be assigned to a SIF that is intended to prevent an event that would produce catastrophic community-wide consequences. After a SIL is assigned to each SIF, each SIF is designed to operate within the designed average probability of failure on demand (PFD) that corresponds to the SIL assigned to the SIF. Because a SIF is typically comprised of a collection of instrumented function components (e.g., a logic solver, sensors, and actuators), and each of the instrumented function components have a respective average PFD, which affects the overall average PFD of the SIF, a designer has some flexibility in the way the overall average PFD is achieved. For example, by assuming a set of environmental conditions (e.g., humidity, temperature and pressure) that the instrumented function components will operate under, a designer is able to arrive at an overall average PFD by establishing regimented testing schedule for each of the instrumented function components. Thus, once a SIS is commissioned, a plant engineer is able to estimate the SIL level of a particular SIF as long as the actual maintenance and environmental conditions do not vary from the assumed design conditions. Unfortunately, after a SIS is operational, a plant engineer is unable to determine what the average PFD or SIL levels are for a SIF once actual testing varies from the regimented test schedule. Furthermore, the actual PFD and SIL levels will vary depending upon actual environment conditions, and as a consequence, a plant engineer will face further uncertainty as to what the actual PFD and SIL level is for the SIP. Corresponding reference characters indicate corresponding components throughout the several views of the drawings. In one embodiment, the invention may be characterized as a method for managing a safety instrumented function including a plurality of instrumented function components. The method including the steps of obtaining, from an asset management application, operating information about at least one of the plurality of instrumented function components; determining a probability of failure on demand for the safety instrumented function based at least in part on the operating information; comparing the probability of failure on demand with a designed probability of failure on demand for the safety instrumented function to establish a variance; and managing the plurality of instrumented function components based on the variance. In another embodiment, the invention may be characterized as a system for managing a safety instrumented function including a plurality of instrumented function components. The system includes an asset management application configured to maintain status information relating to the plurality of instrumented function components; and an online safety integrity level application in communication with the asset management application. The online safety integrity level application is configured to receive the status information and calculate a probability of failure on demand for the safety instrumented function based at least in part on the status information. In a further embodiment, the invention may be characterized as a processor readable medium including processor-executable code to generate safety availability information for an instrumented function including a plurality of instrumented function components. The code includes instructions for: obtaining, from an asset management application, operating information about at least one of the plurality of instrumented function components; determining a probability of failure on demand for the instrumented function based at least in part on the operating information; and generating the safety availability information based on the probability of failure on demand. In yet another embodiment, the invention may be characterized as a method (and means for accomplishing the method) for managing a plurality of instrumented function components. The method including the steps of: receiving, from an online safety availability application, operating information about the plurality of instrumented function components; updating, within an asset management database, status information for the plurality of instrumented function components based upon the operating information; and managing the plurality of instrumented function components based on the status information. In one aspect, the present invention is directed to a safety and reliability monitoring system, also referred to herein as a COSIL™ system, which provides historical, real time and predictive probability failures for an online instrumented system, e.g., a safety instrumented system (SIS), based on events which occur during operation and maintenance of the instrumented system. Unlike current approaches for evaluating safety and reliability, which are generally based upon static offline calculations using assumed average conditions over the life cycle of the instrumented system, the present invention according to several embodiments is capable of providing dynamic, online calculations of average probability of failure on demand, instantaneous probability of failure on demand, and safety integrity level (SIL) using actual events (e.g. time of test) in an industrial plant. In some embodiments, the present invention also provides reliability information (e.g., mean time to fail (MTTF)) based on actual events. As a consequence, the inventive COSIL™ system may be employed to provide accurate continuous online status information for an instrumented function, e.g., a safety instrumented function. The term continuous as used herein should not necessarily be construed to mean that calculations are continually performed (i.e., without interruption). The COSIL™ system according to several embodiments, however, does allow a plant engineer to obtain substantially continuous values of PFD, SIL and/or MTTF, if so desired. It should be recognized that the COSIL™ system also allows calculations to be performed at less frequent intervals, e.g., daily, weekly or monthly. Referring first to FIG. 1 shown is a block diagram of an exemplary industrial system 100 in which a COSIL™ system according to one embodiment of the present invention is implemented. As shown, the system 100 includes a programmable device 102 in communication, via a test input 104, with an actuator 108 and a sensor 110 which implement an instrumented function 112, e.g., a safety instrumented function (SIF). Also shown is an environmental input 106 which may be implemented to provide additional input to the COSIL™ module 114. The programmable device 102 may be realized using any one of a variety of devices, which have input/output (I/O) functionality and contain a CPU and memory and (not shown). The programmable device 102 may be, for example and without limitation, an intelligent field device, a safety controller, a programmable logic controller (PLC), a controller, a general purpose computer, a personal digital assistant (PDA) or potentially any other device that includes a processor, memory and input/output capability. The instrumented function 112 represents a specific function executed by the 108 actuator and sensor 110 to achieve or maintain a safe state for a process with respect to a specific event, e.g., a hazardous event. The sensor 110 and actuator 108, also referred to herein as instrumented function components, respectively monitor and react to process conditions in the industrial system 100 in order to help ensure that the instrumented function 112 is carried out on demand. Although one sensor 110 and one actuator 108 are shown for simplicity, it should be recognized that there are potentially multiple actuators and sensors associated with a particular instrumented function, e.g., a particular safety instrumented function (SIF). One of ordinary skill in the art will recognize that there are several varieties of both sensors and actuators. In one embodiment, for example, the sensor 110 is a pressure sensor and the actuator 108 controls a shut off valve. The test input portion 104 in some embodiments is an automated test input unit, that provides test information, e.g., a most recent test time and date, for the actuator 108 and/or sensor 110 to the COSIL™ module 114 without human intervention. In one embodiment, for example, the actuator 108 and sensor 110 are coupled to the programmable device 102 via a communication link. In other embodiments, the test input portion 104 is a keypad or other user interface device, which allows a plant engineer, for example, to provide test information for the actuator 108 and/or sensor 110 to the programmable device 102. Within the programmable device 102 are shown the COSIL™ module 114 and an I/O module 116. The COSIL™ module 114 according to several embodiments is implemented by software that is read from a memory and processed by a CPU (not shown) of the programmable device 102. The COSIL™ module 114 generally comprises processor-executable code (a “COSIL™ program”) specifically designed to calculate, as a function of operating information for the instrumented function components 108, 110, a probability that the instrumented function 112 will fail on demand. As discussed further herein, the COSIL™ program may be created by one of several quantitative risk/reliability analysis (QRA) methodologies including, but not limited to, function block diagram analysis, fault tree analysis, structured text techniques, simple equation methodology, Markov modeling and reliability block diagram methodology. While referring to FIG. 1, simultaneous reference will be made to FIG. 2, which is a flow chart 200 illustrating steps carried out by the COSIL™ module 114 according to several embodiments of the present invention. In operation, the COSIL™ module 114 initially obtains operating information about at least one of the instrumented function components 108, 110 (Step 201). In several embodiments the operating information includes test information that includes for example, a time and date when a test is successfully performed. In the present embodiment, the COSIL™ module 114 receives the operating information via the test information input portion 104. In some embodiments, this test information is saved in a memory of the programmable device 102, which allows the COSIL™ program to calculate an elapsed time between the time of the test and a present (or future) time. In other embodiments, a timer is triggered that tracks the elapsed time between the time of the test and a present time. Although certainly not required, the operating information received at Step 201 may include environmental information, which characterizes the operating environment for the instrumented function components 108, 110 (e.g., humidity, temperature and pressure). In this way, the COSIL™ module 114 is provided with actual environmental conditions for the instrumented function components 108, 110 in the instrumented function 112. It should be recognized that various modes of operation of the COSIL™ module 114 are contemplated in which, for example, only test information is received, only environmental information is received, or both test and environmental information are received at the COSIL™ module 114. It is also further contemplated that in one embodiment, both test and environmental information are received at the COSIL™ module 114, but the COSIL™ module 114 only utilizes either the test or environmental information. Although the COSIL™ module 114 has been described as receiving test data for one of the instrumented function components 108, 110 in the instrumented function 112, it should be recognized that in several embodiments, the COSIL™ module 114 receives test information on an ongoing (e.g., substantially continuous) basis for potentially hundreds of instrumented function components, and is able to establish an elapsed time since a last test for each of the hundreds of instrumented function components. Once the COSIL™ module 114 has received the operating information about at least one of the instrumented function components 108, 110, the COSIL™ module 114 calculates a probability of failure on demand (PFD) for the instrumented function 112 based on the operating information (Step 202). Although operating information for one or more of the instrumented function components 108, 110 may be received at any given time, it should be recognized that the PFD for the instrumented function 112 is calculated as a function of a PFD for each of the instrumented function components 108, 110 that contribute to the availability of the instrumented function 112 on demand. In some embodiments, the probability of failure on demand calculated in Step 202 is an instantaneous probability of failure on demand, which is calculated using the following equation:PFDINST=1−e−λt Eq. (1)where λ is the failure rate for the element measured in a number of failures per unit of time and t is the elapsed time since the last test of the element. The failure rate λ, and hence PFDINST, will be typically be a function of environmental conditions such as temperature, pressure and humidity. In other embodiments, the probability of failure on demand determined in Step 202 is an average probability of failure on demand, which is calculated using the following equation:PFDAVG=1+[(e−λt−1)/λt] Eq. (2)where, again, λ is the failure rate for the element measured in a number of failures per unit of time and t is the elapsed time since the last test of the instrumented function component. In yet other embodiments, the COSIL™ module 114 calculates both PFDINST and PFDAVG for the instrumented function 112. Although the PFD for the instrumented function 112 is calculated as a function of the PFD of each of the instrumented function components 108, 110, it should be recognized that the PFD for each of the instrumented function components 108, 110 need not be calculated. For example, if one of the instrumented function components 108, 110 has failed (i.e., the instrumented function 112 is in a state of degraded operation), in one embodiment the PFD value for the failed instrumented function component is forced to a predefined value (e.g., 1.0). In this way, a PFD for the instrumented function 112 may be calculated even though one of the instrumented function components 108, 110 has failed. For example, assume an instrumented function includes two instrumented function components “a” and “b,” and the instrumented function fails on demand if both “a” and “b” fail on demand. In non-degraded operation, the probability of failure on demand for the instrumented function is a product of the probability that “a” will fail on demand and the probability that “b” will fail on demand (i.e., P=Pa*Pb). If “a” fails a test, however, then Pa is set equal to 1.0, and the probability that the instrumented function will fail on demand during such degraded operation is P=1*Pb=Pb. After a probability of failure on demand is calculated for the instrumented function 112 (Step 202), the probability of failure on demand is compared with a designed probability of failure on demand for the instrumented function to establish a variance (Step 204). In one embodiment, the variance is simply the difference (potentially positive or negative) between the designed probability of failure on demand and the calculated probability of failure on demand. In some embodiments, the designed probability of demand is a designed average probability of failure on demand. As previously discussed, during a design phase of instrumented functions, e.g., safety instrumented functions, a designer typically establishes a test interval period for each instrumented function component in an instrumented function in order to ensure that an average PFD for the instrumented function is maintained below a designed average PFD level. In other embodiments, the designed probability of failure on demand is a designed instantaneous probability of failure on demand, and the actual instantaneous probability of failure on demand calculated in Step 202 is compared with the designed instantaneous probability of failure on demand. Next, after a variance is established, the instrumented function components 108, 110 are managed based upon the variance. In one embodiment for example, an alarm is provided when the calculated probability of failure on demand for an instrumented function exceeds a designed probability of failure on demand. In the embodiments where the calculated PFDAVG is calculated, for example, an alarm is produced when the calculated PFDAVG exceeds the designed average probability of failure on demand. In other embodiments, as described further herein, in addition to alarm feedback, the COSIL™ module 114 provides historical, on-line and predictive reporting of probability of failure on demand values for several instrumented functions. Again, it should be recognized that the COSIL™ system according to several embodiments tracks test information (and in some embodiments environmental conditions) for several instrumented function components within each of the instrumented functions to arrive at a calculated probability of failure on demand for each respective instrumented function. As a consequence of this wealth of information, a plant engineer is provided with many more management options than prior plant management methodologies. For example, it is often advantageous to perform tests, albeit outside of the prescheduled test regimen, on instrumented function components while a portion of a plant process is shut down for repairs. Testing one or more instrumented function components 108, 110 in the instrumented function 112 before their respective scheduled test dates, however, decreases the probability of failure on demand (PFD) and increases the risk reduction factor (RRF) for the associated instrumented function. Because the present invention, according to several embodiments, provides feedback indicating a resulting probability of failure on demand due to the unscheduled testing, a plant engineer is able to manage both the tested instrumented function components in the instrumented function and other instrumented function components that were not tested based upon the unscheduled testing. For example, if the calculated PFDAVG after the unscheduled testing is reduced substantially below a designed average probability of failure on demand, instead of shutting a process down (and losing productivity) to test other instrumented function components according to their designed schedule, a plant engineer may wait, e.g., until a planned shutdown, with the knowledge that the PFDAVG for the instrumented function is still below the designed probability of failure on demand. Thus, the present embodiment allows a plant engineer to take credit for testing in advance of a scheduled test date, and potentially save a substantial amount of money by keeping a process running longer than would otherwise be possible using prior methodologies. Similarly, in one embodiment the present invention allows a plant engineer to establish a risk if testing of an instrumented function component was not performed as scheduled. This is a significant advantage over prior management methodologies, which leave a plant engineer unsure of whether the actual PFDAVG or PFDINST level exceeds a designed PFD level. Furthermore, in several embodiments the present invention allows a plant engineer to take credit for replacement of instrumented function components. Prior methodologies, which merely establish a fixed test schedule to maintain an acceptable PFD and risk reduction factor (RRF), simply do not provide the means for a plant engineer to take into consideration the effects of replacing several instrumented function components at different times. The present invention according to these several embodiments, however, is able to track both replacement of instrumented function components and variances between actual testing and a designed test schedule to allow a plant engineer to take credit for any increased risk reduction factor (RRF). Yet another advantage of some embodiments of the present invention is the ability to establish PFDAVG or PFDINST as a function of environmental conditions including, e.g., temperature, pressure and/or humidity. In these embodiments, a plant engineer may adjust the test interval or environmental conditions to maintain a PFDAVG or PFDINST in response to varying environmental conditions. In contrast, a plant engineer operating under prior management methodologies cannot tell what effect changes in environmental conditions have on the actual average PFD for any instrumented function. As discussed, prior plant management methodologies included a predetermined testing schedule that assumed a set of environmental conditions. In some embodiments, the calculated probability of failure on demand values (i.e., PFDINST and/or PFDAVG), for safety instrumented functions are converted to safety integrity levels. Referring to FIG. 3 for example, shown is a graph depicting the relationship between safety integrity level and probability of failure on demand. As shown, the relationship is determined by the following equation:SIL=−Log(PFD) Eq. (3) Consequently, based on the on-line calculation of the PFDAVG and/or PFDINST, a corresponding PFDAVG and/or SILINST may be calculated as a real number. Thus, a plant engineer is able to monitor calculated SIL values over time and deduce trends based upon the changes in the SIL level over time. For example, if continuous online SIL levels of 3.3, 3.2, and 3.1 have been respectively calculated over three previous months, a plant engineer is able to determine that the SIL level is about to change from a SIL 3 to a SIL 2, and the plant engineer is able to take action to raise or maintain the SIL level. It should be recognized that in the context of a safety system, the present invention in several embodiments is applicable to both PFD/SIL calculations based on continuous (high demand) mode of operation and low demand operation. Although online calculation of average probability of failure on demand PFDAVG for an instrumented function provides a wealth of information heretofore unavailable to a plant engineer, the ability to calculate an instantaneous probability of failure on demand PFDINST provides even more information to a plant engineer. An average probability of failure on demand, for example, does not provide information about the range of probability of failure on demand values that an instrumented function may render during a period that the PFDAVG is determined. Referring next to FIG. 4, shown is a graph depicting the probability of failure on demand for an instrumented function with respect to time for two different test intervals. Shown is a first graph 402 of an instantaneous probability of failure on demand for an instrumented function tested with an interval TI1. Also shown is a second graph 404 of an instantaneous probability of failure on demand for the same instrumented function, which is tested at an interval TI2. Although the test interval TI1 produces an average probability of failure on demand (PFDavgTI1) which is below a designed average probability of failure on demand (Designed PFDavg), there are significant periods of time during which the actual probability of failure on demand exceeds a designed average probability of failure on demand (Designed PFDavg). This graph indicates that a plant engineer without instantaneous PFD information may erroneously be led to believe that the instrumented function is providing a continuous risk reduction factor (RRF), when in fact it is not. By providing instantaneous probability of failure on demand information to a plant engineer, the plant engineer is able to recognize potential problems, e.g., when the instantaneous PFD exceeds a designed maximum, and make adjustments to test intervals and/or environmental conditions to bring the PFD and RRF of the instrumented function into an acceptable range. As shown in FIG. 4, by decreasing the test interval to TI2 for example, the instantaneous probability of failure on demand 404 for the instrumented function at all times is maintained below the designed average probability of failure on demand (Designed PFDavg). Referring next to FIG. 5, shown is an industrial system or plant 500 in which another embodiment of the COSIL™ system is implemented. As shown, coupled to a network 502 are several programmable devices 102A through 102G including a DCS system 102A, a safety controller 102B, two intelligent field devices 102C, 102D coupled by a field bus 520, a programmable logic controller (PLC) 102E, a controller 102F and a control computer 102G. As shown, within each of the programmable devices is a respective COSIL™ module 114A through 114G. Also shown coupled to the network 502 are a system computer 510 and a personal digital assistant 512. In the present embodiment, each of the programmable devices 102A-102G are coupled to instrumented function components (not shown) that implement one or more instrumented functions, e.g., safety instrumented functions. The programmable devices 102A-102G are also coupled via the network 502 to a system computer 510 and a personal digital assistant 512. Although the programmable devices 102A-102G are able to communicate with the system computer 510 and the personal digital assistant (PDA) 512 via the network 502, it should be recognized that the programmable devices 102A-102G do not necessarily communicate with each other. One of ordinary skill in the art will recognize that a variety of network systems may be implemented to provide a communication path between each of the programmable devices 102A-102G and the system computer 510 and/or the personal digital assistant (PDA) 512. A wireless network, for example, may be utilized as part or all of the network 502. In the present embodiment, each of the programmable devices 102A-102G includes a respective COSIL™ module 114A-114G for calculating a PFDINST and/or a PFDAVG for each of their respective instrumented functions. It should be recognized that some of the programmable devices 102A-102G may receive operating information from more than one instrumented function. For example, each of the programmable devices 102A-102G may be associated with more than one instrumented function, and each instrumented function may include more than one instrumented function component. In operation, each programmable device 102A-102G, and hence, each respective COSIL™ module 114A-114G receives operating information, e.g., test and/or environmental information, about its associated instrumented function components, and calculates a probability of failure on demand for the instrumented function associated with the instrumented function components. In this embodiment, the calculated probability of failure on demand for one or more instrumented functions is forwarded via the network 502 to the system computer 510 where it is provided by a reporting application 516 to the display 514. As discussed further herein, information including a designed SIL level, an on-line SIL level and instantaneous PFD as well as deviation lights/alarms may be displayed on the display 514. As previously discussed, the probability of failure on demand may be converted to a SIL level for convenient reporting to a user at the system computer 510 and/or the personal digital assistant 512. One of ordinary skill in the art will recognize that conversion from a probability of failure on demand to a SIL level may be calculated either in the programmable devices 102A-102G (e.g., in the respective COSIL™ modules 114A-114G) or the system computer 510. In one embodiment, calculated probability of failure on demand values for each instrumented function are forwarded to the personal digital assistant (PDA) 512 (e.g., via a wireless link). The personal digital assistant 512 may be any portable computing device with programming and reporting capability including, but not limited to, cellular telephones and notebook computers. The portable aspect of the PDA allows a plant manager to receive alarms and/or generate reports without being “tied” to a desktop-type computer. Referring next to FIG. 6, shown is one embodiment of the safety controller 102B of FIG. 5 in accordance with one embodiment of the present invention. As shown in FIG. 6, the safety controller 602 includes a COSIL™ module 604 located within a control programs portion 606 of the safety controller 602 and is in communication with a tester 608 to receive information about testing of instrumented function components in a plant 600. Also shown is an environmental input, which may be utilized along with the information about testing to calculate an average probability of failure and/or an instantaneous probability of failure on demand for an instrumented function based upon the test and environmental information. In some embodiments, the tester 608 is an operator that inputs test information manually into the safety controller 602, and in other embodiments, the tester 608 is an automated test feedback device that updates the COSIL™ module 604 automatically with any test information. As depicted in FIG. 6, the safety controller 602 provides an alarm 609 to an operator 610 without communicating via the network 502. In one embodiment, for example, the safety controller 602 does not communicate any PFD or SIL information to other devices and simply provides an alarm if any instrumented functions have a PFD level that rises above a designed PFD level. Referring next to FIG. 7, shown is an industrial system 700 in which the COSIL™ system is centrally operated according to one embodiment of the present invention. As shown in FIG. 7, the present embodiment includes a collection of programmable devices 702, 704, 706, 708, 710, 712, 714, which include the same type of programmable devices described with reference to FIG. 5, but in the present embodiment, a system computer 716 calculates PFD information for each of the safety instrumented functions and provides, via a display 718, PFD and/or SIL information for each of the instrumented functions. It should be recognized that each of the programmable devices is associated with an instrumented function (e.g., the instrumented function 112), and each instrumented function includes instrumented function components (e.g., the instrumented function components 108, 110). For clarity, however, the associated instrumented functions and instrumented function components are not shown. Referring briefly to FIG. 7A, shown is the COSIL™ module 720 of FIG. 7 according to one embodiment. As shown, the COSIL™ module 720 includes N separate COSIL™ programs 7221-722N that correspond to N respective instrumented functions in the plant 700. In one embodiment, each of the programmable devices 702, 704, 706, 708, 710, 712, 714 forwards operating information (e.g., test information) to the system computer 716 and/or the PDA 724 about each of the instrumented function components that the programmable device is associated with. In another embodiment, operating information (e.g., test information) about instrumented function components is provided to the system computer 716 by manual entry of a user (e.g., as tests are performed). Each of the COSIL™ programs 7221-722N in the COSIL™ module 720 is associated with a corresponding one of N instrumented functions and tracks operating information for each instrumented function component (e.g., each of the instrumented function components 108, 110) in the corresponding instrumented function (e.g., the instrumented function 112). Based on the operating information, each of the COSIL™ programs 7221-722N calculates, on an ongoing basis, the probability of failure on demand for the corresponding one of the N instrumented functions. In this way, the system computer 716 is able to provide alarms responsive to actual plant events and/or conditions. As discussed herein, the COSIL™ module 720 in some embodiments also includes historical and predictive reporting capabilities in addition to on-line reporting. Referring back to FIG. 7, the COSIL™ module 722 in an exemplary embodiment is implemented in a personal digital assistant (PDA) 724. In this embodiment, the COSIL™ module 722 operates in much the same way as the COSIL™ module 720 in the system computer 716, i.e., the COSIL™ module 722 tracks operating information for each instrumented function component in each instrumented function and calculates, on an ongoing basis, the probability of failure on demand for each monitored instrumented function. In addition, the COSIL™ module 722 may generate alarms and reports for a user, but this is not required. Referring next to FIG. 8, shown is one embodiment of a system computer 800 that may be implemented to carry out the functions of the system computers 510, 716 of FIGS. 5 and 7. As shown, the system computer 800 includes a quantitative risk/reliability analysis (QRA) portion 802, which converts information about each instrumented function into one corresponding COSIL™ program. As discussed, each COSIL™ program (which may be stored in the memory 804, the COSIL™ module 720 of the system computer 716, the COSIL™ module 722 the PDA 724 and/or in the COSIL™ modules 114A-114G of the programmable devices 102A-102G) provides a PFD value for an associated instrumented function (e.g., the instrumented function 112) based on operating information about instrumented function components (e.g., the instrumented function components 108, 110) included in the instrumented function. In an exemplary embodiment, the QRA portion 802 utilizes function block diagram analysis that allows a user to convert an instrumented function fault tree into a function block diagram. The QRA portion 802 then converts the function block diagram into a COSIL™ program for the instrumented function. In one embodiment, the QRA portion 802 is implemented with a Triconex® TS1131 application, but this is certainly not required. In one embodiment, to provide assistance to a user converting a fault tree to a function block diagram, the user is provided with one or more electronic files which include a library of function blocks, e.g. AND and OR logic function blocks, along with Eq. (1) and Eq. (2) set forth above. Such function blocks and equations may be tailored to be read and utilized by various QRA software applications including the Triconex® TS1131 application. In addition, in some embodiments, exemplary function block diagrams are provided to the user to further guide the user. In other embodiments, other QRA methodologies are utilized to create COSIL™ programs for each instrumented function including, but not limited to, structured text techniques, simple equation methodology, Markov modeling and reliability block diagram methodology. It should be recognized that the QRA portion 802 need not be implemented in the system computer 800, and in other embodiments, the COSIL™ programs are created by the user on other machines, or simply provided to the user (e.g., from a third party). In some embodiments (e.g., when the system computer 800 is implemented within the system 700 described with reference to FIG. 7), each COSIL™ program is stored in a memory 804 of the system computer and a CPU carries out the instructions in the COSIL™ program to calculate a PFD for each instrumented function. In these embodiments, an input/output (I/O) portion 806 receives (e.g., from the network 726) operating information for instrumented function components in each instrumented function. In other embodiments (e.g., when the system computer 800 is implemented in the system 500 described with reference to FIG. 5), after a COSIL™ program is created for an instrumented function, it is provided (e.g., uploaded via the network 502), to a programmable device (e.g., one of the programmable devices 102A-102G) where it is stored and carried out by a CPU on the programmable device. In these embodiments, the I/O portion 806 receives PFD and/or SIL information from programmable devices (e.g., the programmable devices 102A-102G) for instrumented functions that are associated with each programmable device. In one embodiment, Foundation Fieldbus function blocks may be uploaded along with the COSIL™ programs to the COSIL™ modules 114C, 114D of the intelligent filed devices 102C, 102D (which are compatible with the Foundation Fieldbus protocol). In yet other embodiments, COSIL™ programs are stored in one or more programmable devices in addition to the system computer 800. Thus, implementations that combine aspects of each of the systems 500, 700 described with reference to FIGS. 5 and 7 are well within the scope of the present invention. Also shown in the system computer 800 is a COSIL™ safety availability application 808 (referred to herein as a COSIL™ application 808). In several embodiments the COSIL™ application 808 includes code to produce a graphical user interface on the display 810, which provides user feedback and user controls (e.g., icons) that allow a user to request several variations of reports for the instrumented functions. For example, information including design SIL levels, continuous PFD and/or SIL levels and instantaneous PFD levels may be displayed for each instrumented function on an ongoing basis. Moreover, alarm information is provided via the display for each instrumented function. In an exemplary embodiment, the COSIL™ application 808 allows a user to analyze historical and future probabilities of failure for each instrumented function in addition to on-line PFD information. Historical operating information for historical analysis may be stored in the memory 804, or may gathered based on retained records (e.g., test records). Beneficially, such historical analysis may be used to reconstruct what the PFD levels were at the time of a prior event. For example, if a plant experienced a boiler explosion, a historical analysis may be performed to determine PFD levels for instrumented functions associated with the boiler. Such historical analysis may provide probative information during an accident investigation of such an event. The COSIL™ application 808 also allows a user to predict future PFD and or SIL levels. For example, a user is able to enter a hypothetical scenario, which includes a future date and a set of assumed conditions (e.g., assumed test intervals and/or environmental conditions). Based upon the information provided by the user, the COSIL™ application 808 calculates PFD and/or SIL values for the instrumented function for the future date based upon the assumed conditions. This functionality allows a plant engineer to test various potential courses of action and make an informed decision based on the results provided by the COSIL™ application 808. Moreover, the COSIL™ application 808 allows future PFD and/or SIL levels to be predicted based upon historical PFD information. Specifically, the COSIL™ application according to one embodiment, tracks and reports PFD and/or SIL level changes for each instrumented function over a period of time. Based upon the tracked information, trends may be established allowing a user to predict when an instrumented function is about to drop below a designed SIL level. As discussed, SIL levels may be reported as real numbers to allow small changes in SIL levels to be perceived by the user. Also shown is an asset management application 812, which according to an exemplary embodiment both receives information from the COSIL™ application 808 and provides information to the COSIL™ application 808. The asset management application 812 may be realized by adapting one of many presently available asset management applications so that it communicates with the COSIL application 808 as described herein. The Avantis™ asset management applications from Invensys® are examples of presently available asset management programs. In an exemplary embodiment, the asset management application 812 tracks replacement of instrumented function components. Specifically, when an instrumented function component is replaced, the asset management application 812 informs the COSIL™ application 808. In this way the COSIL™ application 808 is able to update the COSIL™ program that is associated with the replaced instrumented function component. In turn, the COSIL™ program resets the elapsed time associated with the instrumented function component as though a test were just performed on the replaced instrumented function component. Conversely, when a test is performed on an instrumented function component, the COSIL™ application 808 receives operating information (e.g., test information indicating whether the test was successful or not) and provides the asset management application 812 with the operating information. The asset management application 812 then updates status information for the instrumented function component in an asset management database 814. In this way the asset management application 812 is provided up to date status information for instrumented function components. As a consequence, instrumented function components can be managed based upon information received from the COSIL™ safety availability application 808. For example, procurement of inventory to replace instrumented function components or their constituent parts may be initiated by the asset management application 812 based upon the status information (e.g., based upon status information indicating the instrumented function component has failed). It should be recognized that information may be transferred between the asset management application 812 and the COSIL™ application 808 according to various techniques. For example, each application 808, 812 may be configured to communicate according to the other application's specific application program interface (API). Alternatively, the applications 808, 812 may exchange information according to well-known communication formats (e.g., using extensible markup language (XML)). It should also be recognized that the asset management application 812 may be located remotely from the system computer 800 and communicate with the COSIL™ application 808 via a network connection. On the other hand, one of ordinary skill in the art will appreciate that the asset management application 812 and the COSIL™ application 808 may be bundled, distributed and installed on the system computer 800 as a single application instead of operating as separate discrete (albeit communicatively coupled) applications. While the invention herein disclosed has been described by means of specific embodiments and applications thereof, numerous modifications and variations could be made thereto by those skilled in the art without departing from the scope of the invention set forth in the claims. For example, the present invention is readily adaptable to providing online mean time to failure (MTTF) information for an instrumented function. As one of ordinary skill in the art will appreciate, the quantitative risk/reliability (QRA) methodologies utilized to provide a COSIL™ program may be modified so that the COSIL™ program calculates MTTF values instead of probability of failure on demand (PFD) values. Although testing intervals are typically not part of an MTTF calculation, it is contemplated that operating information including notice of a failure of an instrumented function component will be utilized in such a calculation. Although instrumented function components are typically replaced quickly upon failure, knowledge of the MTTF value while an instrumented function component is nonfunctional provides a plant engineer with information to make a more informed decision about operating the instrumented function until the instrumented function component is replaced. |
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052162559 | description | DETAILED DESCRIPTION FIG. 1a shows an exemplary embodiment of the beam defining system 6 of a medical linear accelerator in accordance with the present invention. An electron accelerator 104 has an exit window 1, through which an electron beam e.sup.- is transmitted to a collimator assembly 105. The collimator assembly 105 includes a target 3 and electron absorber 9 for generating an X ray beam. The generated beam is substantially free of unabsorbed electrons. The target 3 and electron absorber 9 are mounted within a carrier plate 8 of the assembly 105. The target 3, absorber 9 and carrier plate 8 may be removed from the apparatus if desired, to use the electron beam itself for treatment (e.g., for superficial treatment) instead of X rays. The term radiation beam will be used to refer to either an electron beam or an X ray beam. The intensity of the electron beam and thus of the X ray beam is controlled by the processor 21. In response to address values provided by the processor 21, a read-only memory (ROM) 100 applies programmed digital values to a digital-to-analog converter (DAC) 102. The DAC 102 converts the digital values into control voltage signals for a modulator 103. The modulator, in turn, controls a pulse injector 106 to provide pulses at a rate determined by the control voltage signals to an electron gun 104a of the linear accelerator 104. In response to different control voltage signal values, the linear accelerator produces bursts of electrons at respectively different pulse rates (commonly referred to as PRF). Bursts produced at a relatively high rate generate a more intense beam of radiation (i.e. a higher dose rate) than bursts at a relatively low rate. In the exemplary embodiment of the invention, the processor 21 can change the intensity of the X ray beam during a treatment to generate different beam profiles. The processor 21 monitors the intensity of the beam using a radiation detector 17. This detector 17 may be, for example, a conventional low-density scintillation detector. The radiation beam has a central move axis 11a. A beam collimating block 4 is disposed in the path of the radiation, directly below the carrying plate 8. The beam collimating block includes a thick walled collimator shielding block or collimator 10. The collimator 10 houses an insert 10a, to which a flattening filter 5 may be mounted. The flattening filter 5, when used, symmetrically attenuates the radiation more towards the center of the beam, so that the intensity of the radiation is approximately constant across the beam width. Filter 5 is rotationally symmetric and is centered relative to axis 11a. The collimator assembly 105 has two pairs of tungsten X ray shielding plates, 12a, 12b, and 13, 14 which are adjustable relative to the axis 11a. In the system shown in FIG. 1a, plate 12a (not visible in the Figure) moves into the page while plate 12b moves out of the page. Plates 13 and 14 move to the left and right respectively. Thus, each pair of plates 12a, 12b and 13, 14 is movable along a single axis, referred to as the X and Y axes, respectively. The X and Y axes and the beam axis 11a form an orthogonal set. The inner edges of the plates define the radiation field edge, and therefore, the positions of the four plates determine the radiation field size. The plates 12a, 12b, 13 and 14 are shown in a view along the Z axis in FIG. 1b. At least two of the plates are capable of crossing axis 11a. The collimator is mounted for rotation about the beam axis. In order to provide the desired accuracy for the speed and position of the collimator plates 12, 13, 14, the plate positions are controlled by an automatic drive units 19 and 20 under computer 21 control. Although only two drive units are shown, it is contemplated that each of the plates 12a, 12b, 13 and 14 may have a separate drive unit, each independently controlled by the processor 21. In this configuration, the processor 21 can independently control the position, velocity and acceleration of each plate during a treatment. The drive units 19 and 20 may be, for example, a conventional numerically controlled servo system which may use either conventional servo motors or stepper motors to control the positions of the jaws 12a, 12b, 13 and 14. Feedback on the instantaneous position of the plates 12b and 13 is provided by respective position sensors 7a and 7b. The sensors 7 and 9 may be, for example, conventional resolver units, mounted on the same shaft as the motors of the drive units 20 and 19, respectively. In the exemplary embodiment of the invention, the computer 21 periodically calculates a desired position for at least one of the jaws 12a, 12b, 13 and 14 and applies the desired positions to the actuators 19 and 20 via actuator control signals. The actuators, in turn, move the jaws to the desired next position. Using this control scheme, a wide variety of beam profiles may be generated which employ both linear and non-linear jaw-motion functions. In addition, the processor 21 may transmit rotation control signals to cause the entire collimator assembly 105 except for the collimating block 4 to rotate by 90.degree. (i.e. counter-clockwise out of the page) and then back to 0.degree. by activating drive unit 22. The invention includes an apparatus for producing a beam with an arbitrary two dimensional isodose contour. An isodose contour is the locus of points in three dimensional space which receive the same total dosage of radiation. The isodose contour is the three dimensional analog of the two dimensional isodose curve. In some of the embodiments of the invention described below, the apparatus used is as described above, with the flattening filter installed in the collimator. With this filter in place, the beam leaving the collimator is of substantially uniform intensity in both the X and Y directions. Other embodiments of the invention produce a substantially flat beam profile without using a flattening filter. In order to determine the plate movements which result in the desired dosage being applied, a coordinate system is adopted in which the X and Y axes are located in a plane parallel to the surface of the object which is to be irradiated. The Z axis coincides with the longitudinal axis of the beam and the positive Z direction is the direction of the beam (i.e., pointing from the radiation source towards the treatment area). One set of plates moves in a direction parallel to the X axis and the other set of plates moves in the direction parallel to the Y axis. The inventor has determined that for an arbitrary dosage distribution, the plate movements and dosage rate are related by equation (1). ##EQU1## where z(x,y)=Depth of isodose contour at (x,y), measured in the z direction D.sub.0(x,y) =Dose rate deposited at the surface at point (x,y) .mu.=Liner attenuation coefficient for the medium irradiated D.sub.a =Dosage on isodose contour v.sub.x =dx/dt=relative plate velocity in the X direction and v.sub.y =dy/dt=relative plate velocity in the Y direction For an isodose contour with an arbitrary shape, the contour z.sub.(x,y) will be a function of both X and Y. In order to apply equation (1) for such an isodose contour, the irradiated surface is divided into a two dimensional array of treatment areas, where an independent radiation field is applied to each area. To apply the radiation to one of these areas, one set of collimator plates 13, 14 is held still, while the plates 12 in the second pair may be moved relative to one another to produce, for example, a wedge shaped area isodose contour. For any area with a flat isodose contour, both sets of plates are held still. For each of these areas, a dosage profile (e.g., constant or wedge shaped) is applied, to approximate the desired isodose contour with a function which is piecewise continuous. This dosage profile may have discontinuities in its derivative at the edges of each treatment area, depending on the profile within each area. The dosage profile may also be changed by changing the intensity of the beam provided by the linear accelerator 104 and wave guide 2. As set forth above, this occurs when the processor 21 changes the address value applied to the ROM 100, thereby changing the PRF of the bursts applied to accelerator 104. The method described above for an arbitrary isodose contour may be time consuming if the number of treatment areas is very large. Depending on the nature of the isodose profile in each area, the collimator plates may have to be repositioned when each treatment area is begun. The method is useful, however if extremely tight control of the isodose contour is desired. The first exemplary embodiment of the invention includes a method for generating a large and useful class of isodose contours for which the number of independent treatment areas is one. That is, the radiation may be applied in a two part treatment consisting of only one set of plate movements in the X direction and one set of movements in the Y direction. While the first pair of plates is moving in the X direction, the plates oriented parallel to the Y axis remain still (i.e., Y=a constant). Similarly, while the second pair of plates is moving in the Y direction, the plates oriented parallel to the X axis remain still (i.e., X=a constant). Further, to simplify the demands made on the equipment configuration, the plate motions are limited so that the plate speed and the beam intensity variations are continuous functions of time within each of the two sections of the treatment. Any contour which can be described by equation (2) falls into this category. EQU z(x,y)=z.sub.1 (x)+z.sub.2 (y) (2) where: z.sub.1 (x)=a function of x only z.sub.2 (y)=a function of y only For any desired isodose contour which can be expressed as the sum of a function of only X plus a function of only Y, the treatment can be applied in two distinct parts, one including plate motion in the X direction for a fixed Y direction plate opening, and the other including motion in the Y direction for a fixed X direction plate opening. An example of such a contour is one in which there is rotational symmetry about the beam axis. For such an isodose contour, any cross section which is perpendicular to the beam axis (i.e., constant depth, z) will be a circle. Such an isodose contour is described by equation (3), in which the locus of points for any fixed value of z define a circle. EQU z(x,y)=C.sub.3 -a * (x.sup.2 +y.sup.2) for all x,y (3) In equation (3), the values of C.sub.3 and "a" are determined from the boundary conditions for a particular isodose contour. In the exemplary embodiment, C.sub.3 is the depth of the isodose curve at beam axis (X=0, Y=0). In order to apply the treatment, the apparatus is initially set up with flattening filter 5 in place, and the lower collimator plates 13, 14 positioned at edges of the beam field symmetrically placed about the X axis at coordinates (X=0, Y=-Y.sub.0) and (X=0, Y=+Y.sub.0), respectively, where Y.sub.0 is a constant. These plates are held steady during the first section of the treatment, and for the purposes of the treatment, may be considered "fully open." While plates 13 and 14 may be physically capable of opening further, this would result in undesirable irradiation of surrounding tissues. Upper plate 12a is initially placed along the X axis at (X=-X.sub.0, Y=0), where X.sub.0 is a constant. Upper plate 12b is placed at the origin (i.e. X=0 and Y=0). Plate 12b is held motionless at the origin, while Plate 12a is moved towards plate 12b at a constant velocity, v.sub.x. It is understood that the application of a radiation beam with constant intensity during this motion would result in the known wedge shaped isodose contour. Instead, during this portion of the treatment, the intensity of the beam is controlled by the processor 21 as a function of the position of plate 12a according to equation (4). ##EQU2## The inventor has determined that this combination of plate motion and radiation intensity produces an isodose contour with the desired parabolic cross section, with the maximum radiation dosage at the origin, and zero dosage at (X=-X.sub.0). Once the two plates meet at the origin, the beam is completely blocked. The irradiation of the half of the treatment area for which X is less than zero is complete for the first half (with Y=a constant) of the treatment. The treatment may optionally be interrupted at this point with no effect on the total dosage received at any point. During the next portion of the irradiation, plate 12a is held motionless, while plate 12b moves away from plate 12a in the positive X direction. The intensity of the radiation beam during this part of the treatment is again controlled to follow equation (4). The second portion of the irradiation treatment deposits a beam profile on the positive side of the X axis, completing the desired parabolic cross section, with the maximum radiation dosage at the origin, and zero dosage at (X=+X.sub.0). When plate 12b reaches the point (X=+X.sub.0, Y=0), the radiation is interrupted by, for example, conditioning the accelerator 104 to provide no electron beam pulses or by closing a shutter to block the electron beam e.sup.-. The first half of the treatment, in which the positions of plates 13, and 14 are held constant, is complete. To set up for the second half of the treatment, the upper plate 12a is returned to (X=-X.sub.0, Y=0), "fully opening" the upper plates 12a and 12b. These plates will be motionless in the second half of the treatment As in the first half of the treatment, the motionless plates are only opened up enough to irradiate the zone to be treated so that surrounding tissues are not subjected to unnecessary irradiation. Lower plate 14 is moved to the origin and lower plate 13 is initially left at its open position at (X=0, Y=-Y0). Plate 14 is held motionless at the origin, while Plate 13 is actuated towards plate 14 at a constant velocity v.sub.y. During this portion of the treatment, the intensity of the beam is governed by equation (5). ##EQU3## This isodose contour has the desired parabolic cross section, with the maximum radiation dosage at the origin, and zero dosage at (Y=-Y.sub.0). Once the two plates meet at the origin, the beam is completely blocked. The irradiation of the half of the treatment area for which Y is less than zero is complete for the second half of the treatment (with X=a constant). During the last portion of the irradiation, plate 13 is held motionless at the origin, while plate 14 moves away from plate 13 in the positive Y direction. The beam intensity during this part of the treatment is again controlled to follow equation (5). The last portion of the irradiation deposits a beam profile on the positive side of the Y axis, completing the desired parabolic cross section, with the maximum radiation dosage at the origin, and zero dosage at (Y=+Y.sub.0). When plate 14 reaches the point (X=0, Y=+Y.sub.0), the radiation is interrupted. The treatment is complete. In these exemplary treatments, the processor 21 moves the plates 12a, 12b, 13 and 14 with fixed velocities and periodically determines desirable radiation beam intensities according to these equations. Radiation intensity is changed by controlling the rate at which electron pulses are emitted by the electron gun. During each of the described treatment schemes, only one of the plates 12a, 12b, 13 and 14 is in motion at any given time. Alternatively, the intensity of the radiation beam may be held constant or allowed to vary in time according to a predetermined function during the treatment and the plates may be moved with velocities that are functions of time or of radiation intensity to produce a non-linear beam profile. To generate a rotationally symmetric beam profile, for example, the equations 4 and 5 may be solved for v.sub.x and v.sub.y and the beam intensity may be held constant. In this alternative embodiment, the processor 21 causes motor controller 18 and drive units 19 and 20 to periodically move the plates to positions which produce the desired velocity profile. A second embodiment of the invention for depositing the desired parabolic isodose profile may be used to overcome hardware limitations on the radiation treatment apparatus which restricts the motion of the lower plates. In the second method, the lower plates may have limited ability to move, or they may even be fixed. In this embodiment of the invention, the first interval of the composite treatment is performed with the lower plates 13, 14 fixed, while the intensity of the radiation beam and the motion of plates 12a, 12b defined by equation (4), as in the first embodiment. At the completion of the first half of the treatment, the radiation is interrupted, the upper plates 12a, 12b are returned to their original positions, and the lower plates 13, 14 remain open. The lower part of collimator assembly 105 is rotated ninety degrees by the drive unit 22 responsive to the processor 21, so that the upper plates 12a, 12b are positioned along the Y axis, and the lower plates are symmetrically placed about the X axis. The second interval of the composite treatment follows the beam intensities and plate motions governed by equation (5), as in the first embodiment. In this second half of the treatment, however, it is the plates 12a and 12b which are moved. The second embodiment may have advantages over the first embodiment for a beam forming apparatus with a rotatable collimator. Because the upper plates are closer to the radiation source, movement of an upper plate effects a greater change in the width of the beam (in the plane of the treatment area) than does movement of a lower plate through the same distance. Since the upper 12a, 12b and lower 13, 14 plates are typically actuated by the same type of equipment, they are each capable of being actuated at the same maximum plate velocity. Therefore, the upper plates 12a, 12b are capable of increasing or decreasing the width of the beam by a desired amount faster than the lower plates 13, 14 can. In addition, control of the apparatus may be simplified in a system having a rotating collimating assembly, since only one pair of numerically controlled drive units 19 is needed to actuate the one set of plates for this device. It is understood by practitioners in the field that the parabolic isodose profiles realized by the above control schemes are exemplary in nature and that a number of variations are mathematically possible. The use of these parabolic contours, rotationally symmetric about the origin simplifies the plate movements used to achieve the desired isodose contours. For example, one or both of the parabolic contours could be offset from the origin by a constant displacement. This would require both plates in one or both sets of plates to be capable of crossing the axis. Although this is technically feasible, it is easier to move the object treated and keep the beam center at the origin than to electronically offset the beam profile center. A third embodiment uses the general teachings of the earlier described embodiments to extend the capabilities of the invention even further. In this embodiment, the collimator 105 does not require a flattening filter 5. Instead, the motions of the collimator plates 12a, 12b, 13, 14 are controlled to produce a beam profile whose isodose contours are approximately flat. That is to say, any isodose contour will lie in a plane parallel to the treatment surface. Using this method, the radiation output of the electron accelerator is not decreased by filter attenuation, so that for a given accelerator, higher radiation intensity may be applied to the treatment area. This method may have many applications ranging from pencil beam treatment to whole body radiation. FIG. 4 shows the beam profile 100 produced with both collimator plates open, when the flattening filter 5 is removed (hereinafter referred to as the "raw beam" profile). The raw beam profile 100 may have an arbitrary form, and will vary with the apparatus used. This profile is empirically determined. For typical raw beams, this isodose curve may be approximated by finding a best-fit parabolic curve 102. The parabolic curve 102 is exaggerated in the Figure to clearly distinguish it from the raw beam profile 100. In order to apply a uniform dosage with the raw beam, it is necessary to expose the areas further from the origin to the beam longer than the center is exposed to the beam. That is to say, a compensating beam profile which is complementary to the parabolic isodose curve of raw beam 100 is needed. FIG. 5a shows how a compensating isodose curve 110 is added to the raw beam isodose curve 102' to provide a flat profile 114. FIG. 5 shows clearly that the compensating isodose curve 110 is greatest at the edges of the beam 118, and has a minimum at the beam axis 116. The smallest total flat beam isodose profile which may be achieved is the dose 120 detected at the beam axis 11a for the raw beam 102'. Mathematically, the compensating isodose profile 110 is the difference produced by subtracting the profile of the raw beam 102' from that of the flat beam 114. To produce this compensating beam profile 110, the compensating dosage profile at the edge of the beam is desirably the raw beam maximum value 120, and the compensating dosage must fall off to zero at the center 11a of the beam. The upper dose limit is determined by the maximum raw beam intensity 120 and the available plate speeds. It is understood by one skilled in the art that a flat profile with a higher total dosage 114' may be achieved by extending the amount of time that the plates spend in any one position. The inventors have determined that the relationship between the plate movements and the dosage rate for the dynamic generation of a flat beam profile is defined by equations (6) and (7). ##EQU4## As noted above, in order to move the collimator plates to furnish the beam profile defined by equations (6) and (7), the treatment area must be exposed to the edges of the beam longer than the center of the treatment area (at the beam axis). As shown in FIGS. 2a and 2b, during any dynamic beam forming treatment, the region under the stationary plate is exposed to the radiation the longest. In contrast to the parabolic beamform generation in the first and second embodiments of the invention, flat beam generation shown in FIG. 5 requires that the stationary plate be positioned at the outer edge of the beam instead of at the beam center. In order to accomplish this, both plates in at least one set of collimator plates 12a, 12b are capable of crossing the axis. In order to apply the treatment, the apparatus is first set up with the lower collimator plates 13, 14 positioned at edges of the beam field symmetrically placed about the X axis at coordinates (X=0, Y=-Y.sub.0) and (X=0, Y=+Y.sub.0), respectively, where Y.sub.0 is a constant. These plates are held steady during the first section of the treatment. As shown in FIG. 5b, upper plates 12a and 12b are both initially placed along the X axis at (X=-X.sub.0, Y=0). Plate 12a is held motionless at the edge of the desired beam, while Plate 12b is actuated towards the origin, increasing the width of the beam. During this portion of the treatment, the intensity of the beam is governed by equation (6). Once plate 12b reaches the origin, the treatment is interrupted. The irradiation of the portion of the treatment area for which X is less than zero and Y is a constant is complete. Next, as shown in FIG. 5c, plate 12a is positioned at the origin and plate 12b is moved to (X=+X.sub.0, Y=0) before beginning the next portion of the treatment. During the next portion of the irradiation, plate 12b is held motionless, while plate 12a moves towards 12b in the positive X direction, decreasing the beam width. The beam intensity during this part of the treatment is again controlled to follow equation (6). The second portion of the irradiation deposits a beam profile on the positive side of the X axis, completing the desired flat beam. When plate 12b reaches the point (X=+X.sub.0, Y=0), the plates have closed and the radiation is interrupted. The first half of the treatment, in which the positions of plates 13, and 14 are held constant, is complete. It should be noted that both plates 12a and 12b cross the beam axis during the first half of the treatment. The upper plate 12a is returned to (X=-X.sub.0, Y=0), "fully opening" the upper plates. As in the first half of the treatment, the motionless plates are only opened up enough to irradiate the zone to be treated, and surrounding tissues are not subjected to irradiation. Lower plates 13 and 14 are both moved to (X=0, Y=-Y0) a closed position, prior to beginning the second half of the irradiation. From this point on, the half of the treatment with the upper plates fixed proceeds similarly to the treatment with the lower plates still. Plate 13 is held motionless at the edge of the beam, while Plate 14 is actuated away from plate 13. During this portion of the treatment, the intensity of the beam is governed by equation (7). The isodose contour has the desired flat profile. Once the plate 14 reaches the origin, the treatment is interrupted. The irradiation of the half of the treatment area for which Y is less than zero and X is constant is complete. Plate 13 is moved to the origin and plate 14 is moved to the edge of the beam at (X=+X.sub.0, Y=0) before resuming treatment. During the last portion of the irradiation, plate 14 is held motionless, while plate 13 moves towards plate 13 in the positive Y direction. The beam intensity during this part of the treatment is again controlled to follow equation (7). The last portion of the irradiation deposits a flat beam profile on the positive side of the Y axis, completing the treatment. The result is that the isodose contour has the desired flat profile. It is understood by one skilled in the art that a flat profile with a higher total dosage 114' may be achieved by reducing the velocities v.sub.x and v.sub.y resulting in the plates being held in each position for a longer period of time. Furthermore, it should be understood by those skilled in the art that other movements of the aperture plates could produce the same result. For example, in FIG. 5c, both plates 12a and 12b could start at position 0, +X.sub.0 and plate 12a could be moved away from plate 12b. It is understood by one skilled in the art that a flat profile with a higher total dosage 114' may be achieved by reducing the velocities v.sub.x and v.sub.y resulting in the plates being held in each position for a longer period of time. It is understood by one skilled in the art that many variations of the embodiments described herein are contemplated. While the invention has been described in terms of exemplary embodiments, it is contemplated that it may be practiced as outlined above with modifications within the spirit and scope of the appended claims. |
abstract | An x-ray window including a support frame with a perimeter and an aperture. A plurality of ribs can extend across the aperture of the support frame and can be supported or carried by the support frame. Openings exist between ribs to allow transmission of x-rays through such openings with no attenuation of x-rays by the ribs. A film can be disposed over and span the ribs and openings. The ribs can have at least two different cross-sectional sizes including at least one larger sized rib with a cross-sectional area that is at least 5% larger than a cross-sectional area of at least one smaller sized rib. |
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claims | 1. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and a thickness thereof in an axial direction differing in a rotational direction; anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of a rotational angle of the beam energy modulator in the rotational direction, the extraction intensity being controlled to vary among different levels on the basis of the rotational angle of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 2. The charged particle beam irradiation system according to claim 1, further comprising:an angle detector for detecting the rotational angle of the beam energy modulator,wherein the controller controls the extraction intensity of the charged particle beam extracted from the synchrotron on the basis of the rotational angle detected by the angle detector. 3. The charged particle beam irradiation system according to claim 1, further comprising:a storage device for storing a plurality of extraction intensity set values of charged particle beams extracted from the synchrotron, the extraction intensity set values corresponding to a plurality of rotational angles of the beam energy modulator,wherein the controller controls the extraction intensity of the charged particle beam on the basis of one of the extraction intensity set values selected from the storage device according to the rotational angle of the beam energy modulator. 4. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and a thickness thereof in an axial direction differing in a rotational direction; anda controller for controlling the amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of a rotational angle of the beam energy modulator in the rotational direction, the amplitude being controlled to vary among different levels on the basis of the rotational angle of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 5. The charged particle beam irradiation system according to claim 4, further comprising:an angle detector for detecting the rotational angle of the beam energy modulator,wherein the controller controls the amplitude of the radio frequency signal on the basis of the rotational angle detected by the angle detector. 6. The charged particle beam irradiation system according to claim 1, wherein the thickness of the beam energy modulator in the axial direction changes in the rotational direction due to a structure provided with a plurality of steps. 7. The charged particle beam irradiation system according to claim 1, wherein the beam energy modulator includes a blade, the thickness of the blade in the axial direction differing in the rotational direction. 8. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and a thickness thereof in an axial direction differing in a rotational direction; anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the thickness of the beam energy modulator in the axial direction, the extraction intensity being controlled to vary among different levels on the basis of the thickness of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 9. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and a thickness thereof in an axial direction differing in a rotational direction; anda controller for controlling the amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the thickness of the beam energy modulator in the axial direction, the amplitude being controlled to vary among different levels on the basis of the thickness of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 10. The charged particle beam irradiation system according to claim 8, wherein the thickness of the beam energy modulator is determined on the basis of the rotational angle of the beam energy modulator. 11. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and a thickness thereof in an axial direction differing in a rotational direction; anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron while the charged particle beam is being extracted, when an ion beam having energy different from the energy to which the shape of the beam energy modulator is optimized is incident on the beam energy modulator, the extraction intensity being controlled to vary among different levels by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device synchrotron on the basis of one of the rotational angle in the rotational direction or the thickness in the axial direction of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 12. The charged particle beam irradiation system according to claim 1, wherein the controller further controls the period for extracting the charged particle beam by starting and stopping the beam extraction on the basis of the rotational angle of the beam energy modulator. 13. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and a thickness thereof in an axial direction differing in a rotational direction; anda controller for outputting an extraction start signal and an extraction stop signal for starting and stopping the extraction of a charged particle beam orbiting the synchrotron and a control signal for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotational angle of the beam energy modulator in the rotational direction, the extraction intensity being controlled to vary among different levels on the basis of the rotational angle of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 14. The charged particle beam irradiation system according to claim 7, wherein, when an ion beam having energy that is lower than the energy to which the shape of the beam energy modulator is optimized is transmitted to pass though the beam energy modulator, the controller reduces the extraction intensity of the ion beam passing through an area where the blade thickness is small compared to the extraction intensity of the ion beam passing through other areas. 15. The charged particle beam irradiation system according to claim 7, wherein, when an ion beam having energy that is higher than the energy to which the shape of the beam energy modulator is optimized is transmitted to pass through the beam energy modulator, the controller reduces the extraction intensity of the ion beam passing through a flat section of the blade of the beam energy modulator where the thickness is small compared to the extraction intensity of the ion beam passing through a flat section of the blade of the beam energy modulator where the thickness is great. 16. A method of extracting a charged particle beam, comprising the steps of:accelerating a charged particle beam using a synchrotron, the synchrotron including a radio frequency applying device for beam extraction;transmitting the charged particle beam extracted from the synchrotron to pass a rotating beam energy modulator, a thickness of the beam energy modulator in an axial direction differing in a rotational direction; andcontrolling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotational angle of the beam energy modulator in the rotational direction, the extraction intensity being controlled to vary among different levels on the basis of the rotational angle of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 17. A method of extracting a charged particle beam, comprising the steps of:accelerating a charged particle beam with a synchrotron, the synchrotron including a radio frequency applying device the beam extraction;transmitting the charged particle beam extracted from the synchrotron to pass a rotating beam energy modulator, a thickness of the beam energy modulator in an axial direction differing in a rotational direction; andcontrolling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of a rotating time of the beam energy modulator in the rotational direction, the extraction intensity being controlled to vary among different levels on the basis of the rotating time of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 18. A method of extracting a charged particle beam, comprising the steps of:accelerating a charged particle beam with a synchrotron;modulating the amplitude of a radio frequency signal applied to a charged particle beam on the basis of a rotational angle of a beam energy modulator in a rotational direction while the charged particle beam is being extracted, the amplitude being modulated to vary among different levels on the basis of the rotational angle of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted;applying the modulated radio frequency signal for extracting the charged particle beam from the accelerator; andtransmitting the extracted charged particle beam to pass a rotating beam energy modulator, the thickness of the beam energy modulator in an axial direction differing in the rotational direction. 19. The method of extracting a charged particle beam according to claim 16, wherein the charged particle beam is extracted from the synchrotron on the basis of extraction intensity set values of charged particle beams corresponding to a plurality of rotational angles of the beam energy modulator. 20. The method of extracting a charged particle beam according to claim 16, wherein the extraction intensity of the charged particle beam extracted from the synchrotron is further controlled by starting and stopping beam extraction. 21. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and transported through the beam irradiation apparatus along a beam axis and having passed said beam energy modulator, the beam energy modulator being rotated in a rotational direction around an axis of rotation extending in a direction along said beam axis and a thickness thereof in the axial direction differing in the rotational direction; anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of a rotational angle of the beam energy modulator in the rotational direction, the extraction intensity being controlled to vary among different levels on the basis of the rotational angle of the beam energy modulator during a period after a start of the extraction of the charged particle beam and before a stop of the extraction of the charged particle beam while the charged particle beam is being extracted. 22. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotational angle of the beam energy modulator, the extraction intensity being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus. 23. The charged particle beam irradiation system according to claim 22, further comprising:an angle detector for detecting the rotational angle of the beam energy modulator,wherein the controller controls the extraction intensity of the charged particle beam extracted from the synchrotron on the basis of the rotational angle detected by the angle detector. 24. A charged particle beam irradiation system comprising:an ion source for generating a charged particle beam;a synchrotron for accelerating the charged particle beam extracted from the ion source, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the a radio frequency applying device on the basis of the rotational angle of the beam energy modulator, the extraction intensity being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus. 25. The charged particle beam irradiation system according to claim 22, further comprising:a storage device for storing a plurality of extraction intensity set values of charged particle beams extracted from the synchrotron, the extraction intensity set values corresponding to a plurality of rotational angles of the beam energy modulator,wherein the controller controls the extraction intensity of the charged particle beam on the basis of one of the extraction intensity set values selected from the storage device according to the rotational angle of the beam energy modulator. 26. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); anda controller for controlling the amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotational angle of the beam energy modulator, the amplitude being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus. 27. The charged particle beam irradiation system according to claim 26, further comprising:an angle detector for detecting the rotational angle of the beam energy modulator,wherein the controller controls the amplitude of the radio frequency signal on the basis of the rotational angle detected by the angle detector. 28. The charged particle beam irradiation system according to claim 22, wherein the thickness of the beam energy modulator in the axial direction changes in the rotational direction due to a structure provided with a plurality of steps. 29. The charged particle beam irradiation system according to claim 22, wherein the beam energy modulator includes a blade, the thickness of the blade in the axial direction differing in the rotational direction. 30. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the thickness of the beam energy modulator in the axial direction, the extraction intensity being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus. 31. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); anda controller for controlling the amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the thickness of the beam energy modulator in the axial direction, the amplitude being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus. 32. The charged particle beam irradiation system according to claim 30, wherein the thickness of the beam energy modulator is determined on the basis of the rotational angle of the beam energy modulator. 33. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); anda controller for controlling the extraction intensity of the charged particle beam extracted from the synchrotron while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus, by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device when an ion beam having energy different from the energy to which the shape of the beam energy modulator is optimized is incident on the beam energy modulator. 34. The charged particle beam irradiation system according to claim 22, wherein the controller controls the period for extracting the charged particle beam by starting and stopping the beam extraction on the basis of the rotational angle of the beam energy modulator. 35. A charged particle beam irradiation system comprising:a synchrotron for accelerating a charged particle beam, the synchrotron including a radio frequency applying device for beam extraction;a beam irradiation apparatus having a beam energy modulator and arranged for irradiating an object with the charged particle beam extracted from said synchrotron and having passed said beam energy modulator, the beam energy modulator being rotated and the thickness thereof in the axial direction differing in the rotational direction; andby controlling an amplitude of a signal supplied to the accelerator an extraction start signal and an extraction stop signal for starting and stopping the extraction of a charged particle beam orbiting the synchrotron and a control signal for controlling the extraction intensity of the charged particle beam extracted from the synchrotron before reaching the beam irradiation apparatus by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotational angle of the beam energy modulator. 36. The charged particle beam irradiation system according to claim 29, wherein, when an ion beam having energy that is lower than the energy to which the shape of the beam energy modulator is optimized is transmitted to pass though the beam energy modulator, the controller reduces the extraction intensity of the ion beam passing through an area where the blade thickness is small compared to the extraction intensity of the ion beam passing through other areas. 37. The charged particle beam irradiation system according to claim 29, wherein, when an ion beam having energy that is higher than the energy to which the shape of the beam energy modulator is optimized is transmitted to pass through the beam energy modulator, the controller reduces the extraction intensity of the ion beam passing through a flat section of the blade of the beam energy modulator where the thickness is small compared to the extraction intensity of the ion beam passing through a flat section of the blade of the beam energy modulator where the thickness is great. 38. A method of extracting a charged particle beam, comprising the steps of:accelerating a charged particle beam with a synchrotron, the synchrotron including a radio frequency applying device for beam extraction;transmitting the charged particle beam extracted from the synchrotron to pass a rotating beam energy modulator, the thickness of the beam energy modulator in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); andcontrolling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotational angle of the beam energy modulator, the extraction intensity being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam energy modulator. 39. A method of extracting a charged particle beam, comprising the steps of:accelerating a charged particle beam with a synchrotron, the synchrotron including a radio frequency applying device for beam extraction;transmitting the charged particle beam extracted from the synchrotron to pass a rotating beam energy modulator, the thickness of the beam energy modulator in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP); andcontrolling the extraction intensity of the charged particle beam extracted from the synchrotron by controlling an amplitude of a radio frequency signal supplied to the radio frequency applying device on the basis of the rotating time of the beam energy modulator, the extraction intensity being controlled while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus. 40. A method of extracting a charged particle beam, comprising the steps of:accelerating a charged particle beam with a synchrotron;modulating the amplitude of a radio frequency signal applied to a charged particle beam on the basis of the rotational angle of a beam energy modulator while the charged particle beam is being extracted from the synchrotron before reaching the beam irradiation apparatus;applying the modulated radio frequency signal for extracting the charged particle beam from the synchrotron; andtransmitting the extracted charged particle beam to pass a rotating beam energy modulator, the thickness of the beam energy modulator in the axial direction differing in the rotational direction, the beam energy modulator being used to control a width of a spread-out Bragg peak (SOBP). 41. The method of extracting a charged particle beam according to claim 38, wherein the charged particle beam is extracted from the synchrotron on the basis of extraction intensity set values of charged particle beams corresponding to a plurality of rotational angles of the beam energy modulator. 42. The method of extracting a charged particle beam according to claim 38, wherein the extraction intensity of the charged particle beam extracted from the synchrotron is controlled by starting and stopping beam extraction. |
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abstract | The invention comprises a method and apparatus for directing protons to a tumor, comprising the steps of: (1) holding a patient with a patient support; (2) providing an imaging system comprising: a rotatable unit at least partially surrounding an axial perimeter of the patient support, a translation guide rail, an imaging source attached to the rotatable unit, and an imaging detector attached to the rotatable unit; (3) translating and rotating the imaging source and the imaging detector relative to the patient support using the translation guide rail and the rotatable unit; and (4) providing an attachment section connected: on a first end to a robotic arm positioning system and on a second end to the patient support and the imaging system, the robotic arm positioning system repositioning, relative to a nozzle system linked to the synchrotron, the attachment system supporting the patient support system and the imaging system. |
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039649687 | abstract | A nuclear reactor fuel assembly, wherein at least some of the fuel elements are provided with spacer members. The spacer members are arranged in a helical line on the lateral surface of each of said fuel elements having spacer members. A spacer member is formed as a bunch of wires, with each wire adjoining at least two neighboring wires along the entire length thereof, and with all the wires of a bunch being rigidly interconnected between the planes of contact of said bunch with the fuel elements disposed adjacent the one whereon said bunch of wires is disposed. |
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062360558 | description | DETAILED DESCRIPTION Referring to FIGS. 1 and 2, a preferred embodiment of an irradiation system according to the present invention includes a radiation source 10, a conveyor system 12, radiation shielding material 14 defining a chamber 15 and an intermediate wall 16 of radiation shielding material. Articles carried by article carriers 17 are transported by the conveyor system 12 in a direction indicated by the arrows from a loading area 18 through a target region, generally indicated at 20, to an unloading area 22. The conveyor system 12 includes a process conveyor 24 for transporting articles carried by the article carriers 17 in a given direction through the target region 20. The radiation source 10 preferably is a 10-million-electron-volt linear accelerator having an electron accelerating wave guide that provides an electron beam for irradiating articles transported through the target region 20 by the conveyor system 12 The radiation source 10 is disposed along an approximately horizontal axis 25 inside a loop 26 defined by a portion of the conveyor system 12 and is adapted for scanning the articles being transported through the target region 20 with an electron beam at a given rate in a plane perpendicular to the given direction of transport by the conveyor system 12. The scanning height and the current of the electron beam are adjusted in accordance with the height and radiation absorption characteristics of the articles being scanned. The scanning of the articles by the electron beam is further controlled as described in the above-referenced U.S. Pat. No. 5,396,074. The accelerator is located inside a removable shield and protected from ionizing radiation and ozone by interior walls. In alternative embodiments, the radiation source scans the articles with a type of radiation other than an electron beam, such as X-rays. The conveyor system 12 includes a power-and-free conveyor throughout and, in addition to the process conveyor 24, further includes a load conveyor 28, all three of which are independently powered. The power-and-free conveyor functions as a transport conveyor for transporting the article carriers 17 at a first given speed from the process conveyor 24 through the unloading area 22 and the loading area 18 to the load conveyor 28. The process conveyor 24 transports the articles carriers 17 through the target region 20 at a second given speed that is different than the first given speed at which the article carriers 17 are transported by the transport conveyor. The load conveyor 28 transports the article carriers 17 from the transport conveyor to the process conveyor 24 at a speed that is varied during such transport in such a manner that when the article carriers 17 are positioned on the process conveyor 24 (that) there is a predetermined separation distance between adjacent positioned article carriers 17. When an article carrier 17 is positioned on the process conveyor 24, the load conveyor 28 is transporting the article carriers 17 at the speed of the processor conveyor 24. Such a conveyor system 12 and the operation thereof is described in detail in the above-referenced U.S. Pat. No. 5,396,074. In order to reorient articles for retransportation through the target region 20 so that such articles can be irradiated from opposite sides, upon it being detected that an article carrier 17 carrying such articles is so oriented as to have been transported through the target region 20 only once, such article carrier 17 is diverted onto aireroute conveyor section 30 and then transported by the transport conveyor past a mechanism 32 that reorients the so-oriented article carrier 17 by 180 degrees for said retransportation through the target region 20. Such a reorienting mechanism 32 and means for detecting the orientation of an article carrier 17 are also described in U.S. Pat. No. 5,396,074 to Peck et al. The radiation shielding material 14 includes walls 14A, 14B, 14C, a floor 14D and a ceiling 14E defining the chamber 15 that contains the radiation source 10, the target region 20 and at least the portion of the conveyor system 12 that includes the process conveyor 24, the load conveyor 28 and the adjacent portions of the transport conveyor. Additional walls 14F of radiation shielding material define an angled passageway 36 into the chamber 15 for the conveyor system 12 and shield the loading area 18 and the unloading area 22, which are located outside of the chamber 15, from radiation derived from the radiation source 10. The intermediate wall 16 is positioned within the loop 26 and transverse to the approximately horizontal axis 25 of the radiation source 10. The intermediate wall 16 has an aperture 38 through which the radiation source 10 is disposed. The ceiling section 14E of the radiation shielding material is supported in part by the intermediate wall 16; whereby the underlying chamber 15 may be of a greater area and/or the ceiling section 14E may of a greater span and/or of a greater weight than would be permitted in the absence of such support. Preferably, the radiation shielding material 14A, 14B, 14C, 14D, 14E, 14F (collectively referred to as 14), 16 is primarily concrete because of cost considerations. However, other types of radiation shielding material may be used when space is limited or in view of other requirements, such as steel. In alternative embodiments, some of the radiation shielding material may be concrete and some not. For example, in one alternative embodiment, the intermediate wall 16 is a type of radiation shielding material other than concrete, such as steel, selected in accordance with limited space requirements, while the remainder of the radiation shielding material 14 is concrete. A beam stop 40 is disposed in a recess 42 in the wall 14A of radiation shielding material that is on the opposite side of the target region 20 from the electron beam radiation source 10. The beam stop 40 is made of a material, such as aluminum, that absorbs electrons and converts the energy of the absorbed electrons into photons that are emitted from the beam stop 40. The beam stop 40 is so disposed in the recess 42 that some of the photons emitted from the beam stop 40 toward the radiation source 10 but obliquely thereto are inhibited from entering the chamber 15 by the portion of the radiation shielding material in the wall 14A that defines the recess 42. The recessing of the beam stop 40 reduces the intensity of back scattered photons, thereby decreasing the thickness required for the side walls 14B, the back wall 14C and the ceiling section 14E. This reduces construction costs and shortens the construction schedule. Sections 44 of the transport conveyor portion of the conveyor system 12 are positioned for transporting the article carriers 17 in directions that are transverse to the given direction of transport by the process conveyor 24. The lateral walls 14B of the chamber-defining radiation shielding material are disposed outside the loop 26 adjacent the (these) transversely positioned sections 44 of the conveyor system 12 and portions of the intermediate wall 16 are positioned adjacent the the transversely positioned sections 44 of the conveyor system 12 and across from substantial portions of the lateral walls 14A. The intermediate wall 16 is thereby positioned between the beam stop 40 and the lateral walls 14B so that photons emitted into the chamber 15 from the beam stop 40 are inhibited from impinging upon the lateral walls 14B. The intermediate wall 16 is also positioned between the beam stop 40 and the wall 14C on the opposite side of the chamber 15 from the wall 14A in which the beam stop 40 is recessed so that photons emitted into the chamber 15 from the beam stop 40 are inhibited from impinging upon the opposite wall 14C. As a result, the lateral walls 14B and the opposite wall 14C may be of a lesser thickness of radiation shielding material than would be required in the absence of the intermediate wall 16. The intermediate wall 16 also is positioned for restricting flow throughout the chamber 15 of ozone derived in the target region 20 from the radiation source 10. Accordingly, most of such ozone can be removed from the chamber 15 by exhaust ducts 46 in the chamber 15 disposed above the target region 20. The dimensions of the various components of the radiation shielding material 14 and of the intermediate wall of radiation shielding material 16 are determined by computer-aided modeling in accordance a technique described in a manual entitled "MCNP--A General Monte Carlo Code for Neutron and Photon Transport" published by the Radiation Shielding Information Center, P.O. Box 2008, Oak Ridge, Tenn. 37831. In an alternative embodiment, the loop within which the intermediate wall 14B is positioned is not a closed loop, such as shown in FIG. 1, but instead is an open loop, such as would be formed by elimination of the. reroute conveyor section 30. An article irradiation system in accordance with the present invention provides the advantages of: (a) reducing the volume of concrete required in the ceiling section 14E, thereby reducing the cost and comiplexity of the structure; (b) reducing radiation levels incident on sensitive electrical and mechanical equipment, such as the radiation source 10 and the reorienting mechanism 32, thereby prolonging the life of such equipment; and (c) constrainig ozone production to the vicinity of the process conveyor 24, thereby reducing the quantity of ozone produced and its dispersal throughout the chamber 15 so as to prolong the life of the equipment and reduce the environmental impact of ozone vented to the atmosphere. The advantages specifically stated herein do not necessarily apply to every conceivable embodiment of the present invention. Further, such stated advantages of the present invention are only examples and should not be construed as the only advantages of the present invention. While the above description contains many specificities, these should not be construed as limitations on the scope of the present invention, but rather as examples of the preferred embodiments described herein. Other variations are possible and the scope of the present invention should be determined not by the embodiments described herein but rather by the claims and their legal equivalents. |
claims | 1. A system, comprising:a nuclear reactor chamber comprising an inlet portion, wherein the chamber is a part of a nuclear power plant;at least one container, wherein each respective container of the at least one container contains liquid nitrogen and cold nitrogen vapor and wherein each respective container includes an outlet portion;at least one thermally activated release mechanism, wherein each thermally activated release mechanism of the at least one thermally activated release mechanism is respectively connected between the outlet portion of one of the at least one container and the inlet portion of the nuclear reactor chamber, each thermally activated release mechanism being configured to release the liquid nitrogen from a connected container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber;a central storage container that stores liquid nitrogen, the central storage container being connected to each of the at least one container, wherein the at least one container can each be independently removed from connection with the central storage container, the central storage container being associated with a refilling mechanism configured to automatically refill the central storage container with liquid nitrogen; anda sensor system, the sensor system activating refilling of any first container of the at least one container, with liquid nitrogen from the central storage container when the amount of liquid nitrogen in the first container is below a predetermined threshold. 2. A system as in claim 1, additionally comprising an apparatus configured to produce liquid nitrogen, the apparatus in fluid and gaseous communication with the central storage container. 3. A system as in claim 1, additionally comprising an apparatus configured to produce liquid nitrogen, the apparatus in fluid and gaseous communication with the at least one container. 4. A system comprising:a nuclear reactor chamber comprising an inlet portion, wherein the chamber is a part of a nuclear power plant;at least one container, wherein each respective container of the at least one container contains liquid nitrogen and cold nitrogen vapor and wherein each respective container includes an outlet portion; andat least one thermally activated release mechanism, wherein each thermally activated release mechanism of the at least one thermally activated release mechanism is respectively connected between the outlet portion of one of the at least one container and the inlet portion of the nuclear reactor chamber, each thermally activated release mechanism being configured to release the liquid nitrogen from a connected container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber;wherein a first container of the at least one container contains boron. 5. A method for stopping or preventing explosions and cooling a nuclear reactor chamber, comprising:connecting an inlet portion of the nuclear reactor chamber through a thermally activated release mechanism to an outlet of at least one container that contains liquid nitrogen and cold nitrogen vapor;configuring the thermally activated release mechanism to release the liquid nitrogen from the at least one container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber;connecting a central storage container that stores liquid nitrogen to each of the at least one container, so that the at least one container can each be independently removed from connection with the central storage container; andusing a refilling mechanism configured to automatically refill the central storage container with liquid nitrogen, including using sensor system to activate refilling of any first container of the at least one container with liquid nitrogen from the central storage container when the amount of liquid nitrogen in the first container is below a predetermined threshold. 6. A method as in claim 5, additionally comprising:using an apparatus configured to produce liquid nitrogen, the apparatus in fluid and gaseous communication with the central storage container. 7. A method as in claim 5, additionally comprising:using an apparatus configured to produce liquid nitrogen, the apparatus in fluid and gaseous communication with the at least one container. 8. A method for stopping or preventing explosions and cooling a nuclear reactor chamber, comprising:connecting an inlet portion of the nuclear reactor chamber through a thermally activated release mechanism to an outlet of at least one container that contains liquid nitrogen and cold nitrogen vapor; andconfiguring the thermally activated release mechanism to release the liquid nitrogen from the at least one container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber;wherein a first container of the at least one container contains boron. |
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063079178 | abstract | A soller slit includes a plurality of metal foils and a plurality of spacers. The spacers are laminated alternatively with the metal foils to support one end portions of the metal foils with a space between adjacent metal foils. The other end portions of the metal foils are opened to be unsupported as a free end. When the soller slit is used in an X-ray apparatus, other X-ray optical components, such as monochromator or a specimen to be analyzed, then the soller slit can be arranged in contact with or in the vicinity of the unsupported end portions of the soller slit. That is, it is possible to unify the soller slit and other X-ray optical components in an assembled state. Therefore, a space dedicated to the soller slit becomes unnecessary. Further, since it is possible to shorten a passage of X-rays correspondingly, attenuation of X-rays to be detected by the X-ray detector can be avoided. |
abstract | The modular lower moving system for nuclear fuel handling includes: a lower reactor vessel assembly including nuclear fuel loaded therein; a carrier having a space allowing the lower reactor vessel assembly to be accommodated therein; a rail extending from a reactor area to a fuel handling area; a transfer cart horizontally movable along the rail; a lifting device installed at the transfer cart, movable upward or downward with respect to the transfer cart; and a drive device configured to supply power to the transfer cart and the lifting device. The method of refueling nuclear fuel using the modular lower moving system includes a carrier lifting process, a lower reactor vessel assembly detachment process, a process of accommodating the lower reactor vessel assembly in the carrier, a carrier lowering process, a transfer cart movement process, a nuclear fuel offloading process, and a nuclear fuel loading process. |
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claims | 1. An apparatus comprising:an outer sheathing of a first metal;an electron generator comprising a second metal disposed within the sheathing and forming an emitter;a positive output pin extending from the emitter through the sheathing;an electron charge collector comprising a third metal disposed between the emitter and the sheathing;a negative output pin extending from the collector through the sheathing, spaced and electrically insulated from the positive output pin;a first layer of insulating material positioned between the sheathing and the collector and a second layer of insulating material positioned between the emitter and the collector;wherein the emitter is coupled to a current source, the current source configured to provide current flow to the emitter;wherein the first metal has an atomic number lower than an atomic number of the second metal; andwherein the third metal has an atomic number lower than the atomic number of the second metal. 2. The apparatus recited in claim 1 wherein the current source is a battery. 3. The apparatus recited in claim 1 wherein the current source is a generator. 4. The apparatus recited in claim 1 wherein the current source is located external to a nuclear reactor vessel comprising the apparatus. 5. The apparatus recited in claim 1 wherein the current flow provided by the current source causes the emitter to generate heat. 6. The apparatus recited in claim 1 wherein the second metal is selected from tungsten, platinum, gold, cadmium, and lead. 7. The apparatus recited in claim 1 wherein the first metal and the third metal comprise an inconel alloy or a steel alloy. 8. The apparatus recited in claim 1 wherein the layer of insulation comprises aluminum oxide or magnesium oxide. 9. The apparatus recited in claim 1 wherein the apparatus is positioned adjacent an interior wall of a core of a nuclear reactor for housing at least fuel rods and coolant for generating heat sufficient to liquefy the coolant. 10. The apparatus recited in claim 9 wherein there are a plurality of apparatuses positioned in axial alignment along the length of the core for measuring the relative power distribution in the core. 11. The apparatus recited in claim 9 wherein there are a plurality of apparatuses positioned circumferentially around the interior of the core. 12. The apparatus recited in claim 11 wherein there are a plurality of apparatuses positioned in axial alignment along the length of the core for measuring the relative power distribution in the core. 13. The apparatus recited in claim 1 wherein the apparatus is supported on at least one wall of a fuel assembly cell of a modular fuel rack. 14. The apparatus recited in claim 13 wherein each wall of the fuel assembly cells of the modular fuel rack supports one apparatus. |
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abstract | The present invention relates to methods and systems for 4D ultrafast electron microscopy (UEM)—in situ imaging with ultrafast time resolution in TEM. Single electron imaging is used as a component of the 4D UEM technique to provide high spatial and temporal resolution unavailable using conventional techniques. Other embodiments of the present invention relate to methods and systems for convergent beam UEM, focusing the electron beams onto the specimen to measure structural characteristics in three dimensions as a function of time. Additionally, embodiments provide not only 4D imaging of specimens, but characterization of electron energy, performing time resolved electron energy loss spectroscopy (EELS). |
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description | This invention relates generally to nuclear reactor safety systems and more particularly, to systems and methods for Standby Liquid Control (SLC). A SLC system of a nuclear reactor injects a liquid, e.g. a boron solution, into the reactor vessel when commanded by the nuclear reactor systems or by an operator of the nuclear reactor. The injection process is sufficient to bring the reactor from full power to a sub-critical condition without control rod movement. Nuclear reactor systems (herein, nuclear reactor and reactor are used synonymously) require periodic surveillance be done to make sure the reactor systems are operating correctly. However, surveillance procedures for analog SLC systems (herein, SLC system, SLC instrument, and SLC Logic Processor and SLC are used synonymously) require reactor personnel to manually actuate the SLC system in order to test the system and obtain reports on the operability of the SLC equipment. Automation of the manual surveillance functions for analog SLCs is not easily achieved. Although analog SLC systems provide a very important reactor safety-related function, more may be done to provide an improved SLC system. A method and system is provided for a nuclear reactor safety related application. The method includes executing two forms of a same application-specific logic, one of the two forms implemented as hardware logic, and the other of the two forms implemented as software instructions for execution by microprocessor-based controlling software. Each form of the application-specific logic is executed with a same set of inputs. The method compares a result produced from the execution of the hardware-implemented form to a result produced from the execution of the software-implemented form. When the compared results concur, the controlling software performs actions associated with the concurring results by executing microprocessor-based software. When the compared results fail to concur, the controlling software reports the failure of the compared results to concur to an operator by executing microprocessor-based software, and thereafter places the microprocessor-based software system into an inoperative (INOP) mode. A digital microprocessor-based system is provided for a nuclear reactor safety related application. Included in the system is a microprocessor with memory, hardware, circuitry, and software programming that provides for execution of two forms of a same application-specific logic, and provides for the two forms to be executed with a same set of inputs. The same application-specific logic is implemented in one form as hardware logic, e.g. within a programmable logic device (PLD). The other of the two forms of the application-specific logic is implemented as software instructions for execution by the microprocessor, e.g. a set of software instructions stored within EPROM memory of the microprocessor-based system. The software programming further provides for comparison of a result produced from execution of the one form of the application-specific logic as hardware logic to a result produced from execution of the other form of the application-specific logic as software instructions. When the compared results concur, the software programming provides for the execution of actions associated with the concurring results. When the compared results fail to concur, the software programming provides for the reporting to an operator of the failure of the compared results to concur, thereafter the software programming executing to place the microprocessor-based system into an inoperative (INOP) mode. In an embodiment of the invention, a digital microprocessor-based SLC Logic Processor injects a boron liquid solution into the reactor when commanded by Anticipated Transient Without SCRAM (ATWS) logic units or by an operator (e.g., by key-switch). For example, the SLC initiates the injection process when commanded by at least two out of four ATWS logic units. FIG. 1 is a block diagram exemplifying an embodiment of a digital microprocessor-based SLC logic unit 10. SLC logic unit 10 includes a SLC logic card 16, at least one contact input card 18 (two shown in FIG. 1) connected to SLC logic card 16, at least one relay card 14 (two shown in FIG. 1) connected to SLC logic card 16, at least one optical card 20 (two shown in FIG. 1) connected to SLC logic card 16, and a front panel display 12 connected to SLC logic card 16. SLC logic card 16 performs logic processing and communication functions for the SLC instrument and associated peripheral cards. SLC logic card 16 communicates with the operator via a main control room (MCR) and MCR panel (MCRP) (not shown in FIG. 1), and with external systems (not shown in FIG. 1), and includes the following functionality. SLC logic card 16 includes a microprocessor 34 which executes a microprocessor operating system that executes SLC application logic. In one embodiment, SLC logic card 16 includes complex programmable logic devices (complex PLDs, or just CPLDs, PLD and CPLD being used interchangeably herein). For example, control CPLD 22 for control and decoding and application and CPLD 24 for SLC injection logic are provided. In an exemplary embodiment, SLC logic card 16 includes EPROM memory 26 for non-volatile program storage, RAM memory 28 for read/write memory needs, and non-volatile RAM 30 (NVRAM 30) for storage of application parameters that cannot be lost when power is removed. Such NVRAM parameters might include self-test error codes and counts, a cold boot counter, a warm boot counter, a watchdog timer, and power supply voltage/current settings (if any). EPROM memory 26 and CPLD 24 also include a same application-specific logic, namely SLC injection logic 36. Each of CPLDs 22 and 24 include a status/results register 32. SLC logic card 16 receives Analog Trip Module (ATM) and tank level sensor inputs and performs logic processing to confirm the sensors operating correctly. SLC logic card 16 uses the inputs to halt the injection process upon a low level indication of boron fluid. The automatic testing and detection of ATM tank level sensors exemplify ability of the SLC instrument to automate surveillance of external equipment and report results to the operator while the plant is on or off line. Contact input card 18 provides for MCRP of Local Panel switch (e.g. key-switch) and Low Voltage Switchgear (LSWG) signal inputs to the SLC logic card 16, for example to control CPLD 22. Each contact input card 18 excites the external contacts which are connected to contact input card 18 and electrically isolates and translates the contact status to SLC logic card 16. Relay card 14 provides control of relay outputs, for example, outputs to injection pumps, injection valves, storage tank valves, and overload bypass control. The relays latch to maintain output control if power is removed from the SLC instrument. Two relay cards 14 are used in series, requiring that both card A and card B (of relay cards 14) energize to control a specific output, thus providing protection against accidental and unwanted injection. Optical card 20 provides conversion between optical and electrical signals. Each optical card 20 receives and transmits to external hardware. For example, optical card 20 receives ATWS injection activation and bypass inputs. Optical card 20 communicates messages to the main control room (MCR) operator. Front panel display 12 provides for a single line display of alphanumeric characters. Front panel display 12 uses steady state LED based technology which is stable and has no flicker. Information that is displayed by front panel display 12 includes system status and mode information, as well as results of self tests. SLC logic card 16 provides for self-test and on-line diagnostics capabilities that allow the operator to identify and isolate failures to replaceable hardware modules, including relay card 14, contact input card 18, power supplies (not shown in FIG. 1), analog trip modules (ATMs) (not shown in FIG. 1), and SLC logic card 16 itself. SLC logic card 16 also contains frequency detector circuitry. The frequency detector monitors an optical input link for one of 3 conditions: a) Active state (used by SLC as a Trip or Bypass command), when a particular optical pulse train frequency is detected (1 Mhz in the logic of one example). b) Inactive state (Not Tripped or Not Bypassed command), which is half the frequency of the active state (500 Khz in the logic of one example). C) Fault state, when no valid Active or Inactive state is detected. The self test logic simulates each of the 3 input conditions and monitors the frequency detector output for a response. Control CPLD 22 provides for system monitoring, read/write one of the two relay card 14, and memory decoding. Control CPLD 22 latches and reads relay card 14 status, reads operator pushbutton and switch (e.g. key-switch) position inputs to contact input card 18, and processes contacts status and bypass and ATWS control inputs received via optical card 20. In one embodiment, control CPLD 22 additionally provides to SLC logic card 16 decoding and latching control for the latching relay driver registers of relay card 14, decoding for LED registers of front panel display 12, control to contact input registers for self testing of contact input card 18, and control for self testing the bypass and ATWS input logic. CPLD 24 provides system monitoring, read/write of the other of the two relay card 14 (the first being controlled by CPLD 22), injection logic status registers, local decoding, and SLC injection logic 36. CPLD 24 contains logic control for pump motors, injection valves, storage tank outlet valves, as well as SLC status monitoring, e.g. overload bypass status monitoring, which may be operated to initiate or halt the injection of boron liquid into the reactor. CPLD 24 provides CPLD 24 status registers and relay card driver control for series relays (which provides protection against accidental and unwanted injection). Status/results registers of a CPLD provide results of execution of CPLD logic. Results in the CPLD status/results registers are compared to results obtained by executing same CPLD logic as software to determine concurrence between the hardware-implemented CPLD logic and the software implementation of the CPLD logic. Such comparison for concurrence provides a check for correct operation of both hardware and software. For example, status/results registers 32 provide the results of executing hardware-implemented SLC injection logic 36 on CPLD 24 and are compared to the results generated from executing the same SLC injection logic 36 as software on SLC logic card 16. Same hardware status inputs, e.g. trip data from Anticipated Transient Without SCRAM (ATWS) logic processors, are used for execution of the software-implemented SLC injection logic 36 as for execution of the hardware-implemented (CPLD) SLC injection logic 36. Only when the results concur between execution of software-implemented and hardware-implemented SLC injection logic 36 is the injection process initiated by SLC logic unit 10. When the results do not concur, a CPLD error is flagged, displayed and reported to the operator. A non-concurrence in addition makes the SLC instrument inoperative, e.g. changing the SLC instrument execution mode to an INOP mode. As noted, the injection process is initiated via a concurrence of votes between execution of software-implemented SLC injection logic 36 and hardware-implemented (CPLD) SLC injection logic 36. Additionally, an external signal is provided to the SLC logic card 16 via an input contact (card 18) to notify microprocessor 34 when the operator manually turns a key at the master control room panel (MCRP) (not shown in FIG. 1) signifying a manual command by the operator to initiate the injection process. In discussing the data flow and execution flow of SLC software 38 (to be discussed in the description of FIG. 5), SLC logic unit 10, and thus SLC software 38, has a plurality of operating modes within which SLC software 38 is executed. For example, in one embodiment, modes within which SLC software 38 executes include operating modes controlled by a front panel key-lock switch and a spring loaded key-lock switch either on the Main Control Room Panel (MCRP) or the Local LSWG Panel. The SLC logic card 16 senses key switch positions via the input contact card 18. In the one embodiment, four positions exist on the SLC Logic Processor front panel: 1. NORMAL, 2. INOP/SELFTEST, 3. PRE-OP OVERLD BYPASS TEST, and 4. OVERLOAD BYPASS TESTThese sub-modes are hereafter called the NORMAL, INOP, PRE-OP and BYPASS TEST positions. As shown in FIG. 2, two positions exist for MCRP/Local Panel mode switch 200: position 202, STANDBY, and position 204, and TEST. FIG. 2 shows the MCRP key switch 206 and indicator lamps 208 for mode selection. On the MCRP, position 210, NORMAL, is the return seat of the spring loaded switch, not a mode. In the exemplary embodiment, STANDBY and TEST are considered the primary modes, while NORMAL, INOP, PRE-OP, and BYPASS TEST are considered as sub-modes. Since the focus is on the SLC Logic Processor, and not the Local Panel or Control Room, the emphasis is on the modes controlled at that SLC front panel, with information pertaining to differences of operation with respect to STANDBY and TEST where required. The combination of the four position key switch on the SLC Logic Processor and the two positions of the MCRP and Local Panel make the following modes: Standby/Normal Standby/Pre-Op Standby/Bypass Test Standby/Normal-Inject Standby/Pre-Op-Inject Standby/Bypass Test-Inject Test/Normal Test/Pre-Op Test/Bypass Test Test/INOP Test/Normal-INOP Test/Pre-Op-INOP Test/Bypass Test-INOP FIG. 3 illustrates the modes/sub-modes 300 of an exemplary embodiment of a SLC Logic Processor and the inputs contributing to transitioning to a given mode/sub-mode. Shown in FIG. 3 are the primary modes standby mode 302 and test mode 312. Inputs 318 contribute to transitioning to the standby mode 302 and inputs 320 contribute to transitioning to test mode 312. As shown in FIG. 3, various sub-modes of standby mode 302 and test mode 312 are attained via the actions listed on the connecting arrows of FIG. 3. Standby mode 302 has the sub-modes NORMAL mode 306, PRE-OP mode 308, BYPASS TEST mode 304, with each of these sub-modes 304, 306, and 308 potentially being in an INJECT sub-mode 310. Test mode 312 has the sub-modes NORMAL mode 314, INOP mode 316, PRE-OP mode 308, and BYPASS TEST mode 304, with each of the sub-modes 304, 308 and 314 potentially being in an INOP sub-mode. While the SLC Logic Processor key-switch is in the INOP position, however, all key switch input from the MCRP or Local Panel is disregarded and the main instrument mode remains in TEST. All modes are selectable by the operator, however, automatic out-of-service, ATWS Initiation Trips, or self-test faults can cause the following sub-modes to be entered: INJECT INOP When the SLC logic unit 10 is in the NORMAL, PRE-OP or BYPASS TEST sub-mode (e.g. keyswitch position), the instrument can be placed into both STANDBY mode or TEST mode from both locations (Local Panel and MCRP). STANDBY/NORMAL mode is the normal operating mode for SLC software 38. In all modes but the INOP sub-modes, SLC software 38 sets LED indicators on front panel display 12, e.g. displaying status information and messages via the LEDs, runs background self tests, and sends self test results and status messages to the operator. Background self tests include all SLC hardware tests excepting tests as outlined in the rest of the operator-selectable modes. Background self tests include memory testing (NVRAM, EPROM & RAM), watch-dog timer counter incrementing, power supply tests, A/D converter testing, CPU, PLD, ATM Sensors and Display Self Test Status. Operator selection of a self-test pushbutton has no function except in the INOP sub-modes. Limited test results are displayed on front panel display 12 and are sent as messages to the operator over communication module interface (CIM) communications links. No background tests are performed on relay cards 14 or input contact cards 18 or any equipment connected to them while not in INOP sub-mode. INOP/SELF-TEST mode is an off-line mode of the SLC instrument during which the injection process must not be initiated. If in progress when switching to INOP mode, the injection process will be halted. A change to INOP mode may occur when finding a critical fault during back ground self-testing. Also, INOP/SELF-TEST mode (technically called one of Test/Normal-INOP, Test/Pre-Op-INOP, Test/Bypass Test-INOP) is selected from the main control room panel (MCRP) by placing the pump into a tagout condition and causes simultaneously ‘INOP’ and ‘TEST’ mode states to be true for execution of SLC software 38. Background self tests are suspended when in INOP mode, and are activated through a self-test pushbutton, which when depressed, causes the execution of the background self tests and additional tests. The additional tests include a) control and testing of relay contact interface logic, b) control and testing of specific latch relays on the outputs, c) simulating inputs for an injection mode and validating the responses of the CPLD, d) executing communications loop back tests, e) frequency detector (located on SLC Logic card 16 to convert ATWS inputs from Fiber Optic Interface Card 20 self test, f) watch dog time-out test, g) single line display hardware test. The additional tests exemplify the ability of the SLC instrument to isolate contacts and simulate inputs [item c) above], as well as isolate and control output relays for testing purposes [items a) and b) above]. As stated, inputs are simulated for initiating the injection process with the CPLD results being validated [items c and e]. Input contact card 18 inputs are also isolated and tested by SLC software 38 without causing actual events or state changes to occur. As stated, control and testing of specific latch relays on outputs as well as of relay contact interface logic provides for isolation of and testing of outputs to external equipment. Test results are displayed on front panel display 12 and are sent as messages to the operator over communications links. All equipment controlled by the SLC Logic Processor (e.g. pump start/stop, MBV-0001 open/close, MBV-0005 open/close, MBV-0001 bypass, MBV-0005 bypass) employ two series relays. When testing, one relay is isolated and latched open permitting testing of the other by latching the other closed and open. In similar fashion, each relay may be tested without energizing the equipment controlled by the relay. All SLC logic is isolated via an override logic located in input contact card 18 to permit testing of each contact input. The operator selects a PRE-OP TEST mode from the front panel of the SLC Logic Processor, the PRE-OP TEST mode being a sub-mode of the STANDBY and TEST modes. This mode causes disabling of the overload bypass relay function to allow manual testing of valves with overload protection enabled. When in PRE-OP TEST mode as a sub-mode of STANDBY mode, SLC software 38 remains in STANDBY mode and is prepared to enter INJECT mode if necessary. When in PRE-OP TEST mode as a sub-mode of TEST mode, SLC software 38 remains in TEST mode and is not prepared to enter INJECT mode. Test results may be displayed on front panel display 12 and are sent as messages to the operator over communications links via fiber optic card 20. The operator selects the BYPASS TEST mode from the front panel of the SLC Logic Processor, the BYPASS TEST mode being a sub-mode of the STANDBY and TEST modes. When in the BYPASS TEST mode, SLC software 38 automatically executes tests that check the capability to energize and detect an overload bypass control solenoid for the relays for the valves mentioned in PRE-OP TEST mode above. When in BYPASS TEST mode as a sub-mode of STANDBY mode, SLC software 38 remains in STANDBY mode and is prepared to enter INJECT mode if necessary. Selecting BYPASS TEST mode as a sub-mode of STANDBY mode exemplifies the ability of the SLC instrument to actuate external equipment (in this case, energize and detect an overload bypass control solenoid for relays which operate valves) for testing purposes while the plant is on or off line. When in BYPASS TEST mode as a sub-mode of TEST mode, SLC software 38 remains in TEST mode and is not prepared to enter INJECT mode. Test results are displayed on front panel display 12 and are sent as messages to the operator over communications links via fiber optic card 20. The INJECT mode is entered automatically in the event of receiving an ATWS trip signal or a pump start key-switch from the MCRP, the INJECT mode being a sub-mode of the NORAML, PRE-OP, and BYPASS TEST modes (while these are also considered sub-modes to STANDBY and TEST, the SLC Logic Processor must be in STANDBY mode). After injection is completed, the system automatically returns to the STANDBY mode. In one embodiment, the INJECT mode is entered by the operator manually turning a key at the main control room panel (MCRP). If SLC software 38 detects a critical fault during any mode of operation, the mode changes automatically to INOP mode, irrelevant to the positions of any user selectable switches on SLC logic unit 10 or the MCRP. Critical faults can to be detected during execution of background self tests. Some critical fault messages include BAD RAM, BAD EPROM, BAD CPLD, BAD CPU, POWER UP FAULT, and CRITICAL ATWS FAILURE. Once changing to INOP mode, SLC software 38 halts any progression of the injection process that may be active. Besides detected faults, the following conditions also cause the SLC instrument to become “Automatically Out Of Service”; Pump Trip Coil Not OK Pump Tripping Power Not Available Pump Closing Power Not Available Injection Valve (MBV-0005) Control Power Not Available Storage Tank Outlet Valve (MBV-0001) Control Power Not AvailableShown in FIG. 4 is MCRP pump control 400. Manual INOP can be selected at the MCRP by selecting ‘Tagout’ position 404 from the pump control key switch 402. Likewise a ‘Tagout” of the Injection Valve (MBV-0005) or the Storage Tank Outlet Valve (MBV-0001) causes a ‘Manual Out of Service’ signal, the signal causing the Logic Processor to become inoperative. The TEST mode is entered by the operator selecting TEST at the MCRP or the Local LSWG Panel, or by switching the SLC instrument to INOP/SELF-TEST mode from the front panel of the SLC Logic Unit 10 (actuated by a key switch located on the front panel display 12). Communication tests on the communication links are run while in INOP/SELF-TEST mode. FIG. 5 is a data flow diagram exemplifying an embodiment of SLC software 38 of SLC logic unit 10. In FIG. 5, square boxes, e.g. injection pump 42, represent external entities to SLC software 38. SLC software 38 is divided into generic software groupings as shown by a plurality of rectangular boxes with rounded corners, e.g. relay control 40. One rectangle box status data store 84 is unlike the others, and represents internal data store for SLC software 38. The box-like entities of FIG. 5 are interconnected by data flow arrows. External data flow out from and into SLC software 38 is depicted by arrows connecting to or from square box entities. Data flow within SLC software 38 itself is shown by interconnecting arrows between the generic software groupings, e.g. the arrow connecting control status 62 with relay control 40, the arrow identified as ‘set a relay’. The following describes the data flow and operation of the software groupings of SLC software 38. Operator 66 powers on the SLC instrument and initialization 82 stores pre-determined initialization state data to status data store 84, some of this data being sent to relay control 40 via control status 62. Initialization 82 initializes the hardware of SLC logic unit 10 by initializing hardware registers to predetermined values and by running tests on the hardware. Relay control 40 receives relay status data, e.g. ‘set a relay’ data, from control status 62, and sets the relays for external equipment, e.g. injection pump 42, overload bypass 44, storage tank valve 46, injection valve 50, and sets the TEST and STDBY lamps of MCRP and front panel display 12 as depicted by TEST & STDBY status lights 48 of FIG. 5. Relay card 14, contact input card 18, and front panel display 12 contained in SLC logic unit 10 are controlled via drivers contained in SLC software 38. Relay control 40, LED control 80, local display 72, input status 52 and contact control 68 exemplify areas of data flow requiring custom I/O drivers. Control status 62 receives control status data from status data store 84, processes it, and sends relay set data to relay control 40. Control status 62 also sends control self test data to self test 70. Contact control 68 receives set contact data from self test 70 and returns contact self test data to self test 70. Contact control 68 isolates contacts for self testing. Self test 70 receives control self test data from control status 62, contact self test data from contact control 68, and/or self test start signal from process pushbutton event 78, executes self tests, and sends resulting self test data to status data store 84 as well as self test status data to local display 72. Background self tests include EPROM self test, RAM self test, power supply self test, CPLD self test, CPU self test, display self test, NVRAM (non-volatile RAM) self test, A/D self test, and frequency detector self test. The SLC software 38 contains drivers and testing algorithms that allow individual or a combination of the frequency detectors to be tested in all modes used (½ MHz mode and a 1 MHz mode in one exemplified embodiment). The SLC software 38 engages (on or off) each of the eight frequency detectors and supplies them with the desired frequency (½ MHz and 1 MHz in one embodiment). Additional tests configured to run in the off-line INOP mode are relay card self test, communications self test, contact card self test, and watch dog self test. Relay card self tests are performed when in the INOP mode and testing isolates the relays so that no injection of boron occurs during the testing. Contact card self tests are performed when in the INOP mode and testing isolates the input contacts when setting and reading back values for the contacts. Each CPLD has specific deterministic outputs based upon the inputs. SLC software 38 repeats the CPLD logic to validate that the hardware is operating correctly. Any difference in the results between the two is flagged as a CPLD error. Each frequency detector is tested with a fault identifying the frequency detector and which aspect of the test failed as described above. Set instrument mode 64 receives mode data from operator 66 and sends this data to input status 52 and local display 72. Local display 72 receives display data from set instrument mode 64, self test 70, process pushbutton event 78, and status data store 84 for updating and changing the display on front panel display 12. Process pushbutton event 78 receives data signals, e.g. pressed soft keys data, from operator 66 and sends data (depending on the signal received from operator 66) to local display 72, LED control 80, and self test 70. LED control 80 receives data signals, e.g. trip reset and lamp test, from process pushbutton event 78 and updates LED status data into status data store 84 for updating front panel display 12 via local display 72. Input status 52 collects external data inputs, e.g. operational status of SLC low voltage switch gear (LSWG) equipment 54, ATWS bypass data from bypass unit 56, tank level low status from ATM & SLC tank level sensors 58, and trip data from ATWS logic processors 60. Input status 52 stores input status data into status data store 84. The input trip data from ATWS logic processors 60 is used as input to SLC injection logic 36 of CPLD 24 of FIG. 1. CPLD 24 uses discrete logic, e.g. ‘and’ and ‘or’ gates along with timers to initiate a half trip alarm condition (if determined by the discrete logic). SLC software 38 periodically polls the hardware status, e.g. status/results registers 32 of CPLD 24 of FIG. 1, and upon realizing a half trip alarm condition, executes software-implemented SLC injection logic 36 to validate the hardware status results. If SLC software 38 concurs with the half trip alarm condition of CPLD 24, SLC software 38 then initiates the other half of the trip alarm for a full trip alarm that is stored in status data store 84. Control status 62 begins the injection process upon receiving concurrence of results (from execution of hardware-implemented SLC injection logic 36 in CPLD 24 and execution of software-implemented SLC injection logic 36 in EPROM memory 26) as indicated by retrieving a full trip alarm status from status data store 84. SLC software 38 changes to INJECT mode based upon inputs. Injection is initiated manually by the operator selecting an MCRP pump start, or automatically by receiving two or more ATWS trip signals. An injection is aborted by the operator selecting an MCRP pump stop, or by making any part of the signal inoperable, e.g. a pump tagout, a system critical self test fault, or by switching SLC software 38 to INOP from the MCRP. Once the tank level as signaled by ATMs falls below an ATM set trip point, the injection pump is stopped, the injection is complete, and SLC software 38 is returned to STANDBY mode. Self tests are run to isolate ATM inputs as well as ATWS inputs in order to test correct functioning of the injection process. Two ATM sensor inputs are supplied. The two inputs are averaged by hardware and fed back to a third ATM, all three ATM signals for tank level sensors are acquired to the data store 84, via input status 52. The SLC software 38 averages the two raw ATM inputs and compares them to the hardware averaged ATM input for self testing. Relays are isolated and controlled during this testing to prevent the actual injection of boron. The injection pump and injection process are started manually or stopped manually by the operator at the MCRP. A key-switch at the front panel display 12 allows the pump to be started or stopped while SLC logic unit 10 is in TEST mode. An injection is not initiated or stopped from front panel display 12. All trips and alarms are initiated primarily by a status change of inputs as received at input status 52 and process pushbutton event 78, for example from the LSWG inputs, ATM or ATWS inputs. In one exemplary embodiment of SLC software 38, all inputs are scanned by an executive loop component of SLC software 38. When an input status changes, the executive loop updates the state machine state data store which in turn means that trips or alarms are updated for processing by control status 62. Send status message 74 obtains message data from status data store 84 and sends status messages to external communication interface module (CIM) 76 for communication to the operator at the MCRP. The flow of execution control for SLC software 38 of FIG. 5 may further be described by the use of FIG. 6. FIG. 6 is a block diagram exemplifying an embodiment of SLC software packaging or component architecture. SLC software component architecture 86 shows an example of a high level packaging structure of SLC software 38. FIG. 6 provides more of an operating systems implementation point of view. The architecture components are interconnected via the dependency arrows among architecture components. Dependencies are shown via the dotted lines with arrows, such as a component ‘A’ at the tail of the arrow “depends on” a component ‘B’ at the arrow head. As an example embodiment, SLC software 38 uses a micro-operating system that includes a system clock 112, timers/scheduler 98, and function manager context switcher 100. All normal functions are implemented through an executive loop 90 that, with the use of system clock 112, timers/scheduler 98, and context switcher 100, changes the software machine from one state to another. The SLC system uses a context switcher 100 and a timers/scheduler 98 to assure that the system can perform the main function, which is repetition and concurrence of the basic SLC injection logic 110 for the purpose of injecting boron into the reactor. All other functions are performed to assure that the system remains in a ready state and can perform the function for which the system was designed. Initialization 94 provides for hardware and software initialization. Architecture components that depend on initialization 94 are watchdog timer 92, global data store 96, timers/scheduler 98, context switcher 100, external device I/O 104, and executive loop 90. Initialization component 94 executes upon power up (cold boot) or upon watchdog timer 92 timeout (warm boot) or upon an attempt to execute an illegal address (warm boot). Initialization 94 sets initial values for hardware, e.g. hardware registers, and software, e.g. global data store 96, and runs tests on the hardware. Once the system is initialized, execution control is given to executive loop 90 which runs simple round robin execution of components that depend upon it. When executive loop 90 finds no events scheduled to run (foreground work), executive loop 90 executes a function, typically a background self test of self test 102, and checks the state machine to determine whether the state machine has changed. Changes in the state machine cause the scheduling of tasks to be run by executive loop 90. In one embodiment, watchdog timer 92 times out and interrupts executive loop 90 to cause another system initialization. System clock 112 resets watchdog timer 92. The system clock 112 is instrumental to scheduling all events and context switching of the SLC software 38 and the watchdog timer 92 assures that the system clock is operating correctly. The system clock 112 assurs the correct operation of the SLC software 38, and the watchdog timer assures correct operation indirectly. Self test 102 depends on executive loop 90. Self test 102 includes different types of functions, which are called and run by executive loop 90 depending on the current system mode. Executive loop 90 depends on context switcher 100 to return control to executive loop 90 once an interrupt or scheduled event has been handled by context switcher 100. State machine events, such as mode changes, input signal changes, and self test faults, are detected from executive loop 90. When detecting a mode change, an input signal change, and/or a self test fault, executive loop 90 invokes a state machine event handler function (not shown in FIG. 6, but part of the context switcher 100) to process the detected condition. Appropriate actions for the detected condition include updating global data store 96 and, depending on the change to the state machine, scheduling other events via timers/scheduler 98. Serial I/O 88 generates and sends SLC status messages via communications links to the MCR. Serial I/O 88 depends on global data store 96 to construct a valid message. A timer is scheduled to expire every ½ second to cause serial I/O 88 to be scheduled and invoked to send messages. Timers/scheduler 98 depends on initialization 94 and on SLC injection logic 110. Context switcher 100 depends on timer/scheduler 98 to signal scheduled events that need to be processed. Timer/scheduler 98 has a system clock 112, which tracks overall time in ‘ticks’. Timer/scheduler 98 functions to allow events to be scheduled and deleted. Scheduled events are time critical events, and are prioritized by criticality. An example prioritization is: function time out, state machine input status change, mode change, self test fault, ATWS mitigation initiation present for ½ second, ATWS mitigation initiation override (abort) present for ½ second, toggle flashing lamps/LEDs every ½ second, injection complete ½ second signal delay, send message every ½ second to MCR, etc. In one embodiment, an event is scheduled directly by a function or by context switcher 100. In most cases, the events to be scheduled are timing events. Each event's wait time is predetermined and set by initialization 94 to prevent certain low priority tasks from monopolizing microprocessor 34. Watchdog timer 92 depends on initialization 94 to initiate watchdog timer 92, depends on executive loop 90 to reset watchdog timer 92 upon completion of a specified function, and depends on context switcher 100 to reset watchdog timer 92 upon completion of a function (if not returning control to executive loop 90). Single line display 106 depends on context switcher 100, self test 102, and SLC injection logic to send messages to single line display 106 for displaying. Single line display 106 also depends on context switcher 100 to signal a step display message to scroll through multiple lines of information. LED control 108 depends upon self test 102, global data store 96, SLC injection logic 110 (e.g. to send a message that a valve is open or closed or that a pump has started or stopped), and context switcher 100. LED control 108 controls SLC front panel display 12, with the exception of self test LEDs and mapping functions that map the status of the state machine to an LED data store. External device I/O 104 is the control interface to all external entities, e.g. the square boxes depicted in FIG. 5, with the exception of serial I/O 88 and LED control 108. External device I/O 104 handles control of all relays, contacts, and NVRAM in the SLC hardware. External device I/O 104 depends on self test 102 and initialization 94. SLC injection logic 110 depends on external device I/O 104 to open or close a relay or contact. SLC injection logic 110 is the software-implementation of SLC injection logic 36, the hardware-implementation being the control logic of CPLD 24. When the SLC is operating correctly, SLC injection logic 110 concurs with CPLD hardware results (that the hardware has performed correctly), concurrence including timing requirements and events to be scheduled. If not concurring, a CPLD fault is signaled, and the mode is changed to INOP mode. SLC injection logic permits valves to open and close, pumps to start and stop, and state machine inputs to be read and updated (state machine mode, e.g. INJECT, is changed only upon concurrence between both the software and hardware). SLC injection logic 110 depends on context switcher 100 to notify SLC injection logic 110 to initiate the injection sequence and logic, and to give notification that key events have occurred at specified times. SLC injection logic 110 depends on global data store 96 to supply global variables that may not be available at context switch time. SLC injection logic 110 depends on external device I/O 104 to open or close a relay or contact. Components that depend on SLC injection logic 110 are single line display 106, and timers/scheduler 98 as already described. SLC global data store 96 supplies SLC library routines used by other software and provide global data that is available across context switches. The dependencies upon and for SLC global data store 96 have already been discussed. Self test 102 depends on executive loop 90 in NORMAL, PRE-OP or BYPASS TEST mode, wherein executive loop 90 schedules self test 102 functions. Self test 102 depends on context switcher 100 to cause invocation of the self test 102 functions in all modes. If in INOP/SELF TEST, self test 102 functions are invoked through a user selectable softkey at the SLC front panel 12, otherwise all background functions are scheduled to run when the Executive Loop 90 is not busy. Context switcher 100 is an event handler and when events occur, execution control is passed to context switcher 100, which handles the event, such as running self test 102 functions. Context switcher 100 handles both scheduled and interrupt based events. All background self test 102 functions are suspended when in INOP/SELF TEST mode. The operator selects the self test pushbutton to generate an interrupt (either directly or by schedule), an event that is caught and processed by context switcher 100, which then calls self test 102 functions. Components that depend on self test 102 are led control 108, external device I/O 104, single line display 102, and global data store 96. Components that self test 102 depends upon are global data store 96, executive loop 90, and context switcher 100. Self test 102 tests that run as background self tests when operating in any mode except INOP/SELF TEST mode are EPROM self test, RAM self test, power supply self test, CPLD self test, CPU self test, display self test, NVRAM self test, A/D self test, ATM Sensor test and frequency detector self test. Additional self test 102 tests that run when operating in INOP mode are relay card self test, communications self test, contact card self test, and watch dog self test and, single line display test. Context switcher 100 catches and processes pushbutton events and state machine events. Context switcher 100 either performs an immediate context switch and services the event (depending on what the event is), or adds the event to the scheduler 98 to be handled as soon as the current function is done executing. Without context switching, the software machine simply polls the input status and performs state-machine updates in an endless loop by executive loop 90. Event handlers are responsible for the context switching and there are event handlers for each of the generic categories of pushbutton events (pushbutton events can be both interrupts or scheduled events depending on the made and state of the machine), state machine events, and timer events. Timer events and timer event schedule updating of the event control block (ECB) are handled directly by system clock 112, which is part of timers/scheduler 98. Timer event context switching is handled from the executive loop 90. Context switcher 100 depends on a pushbutton event handler (not shown in FIG. 6), depends on timers/scheduler 98 which indicate that a timer has expired and a scheduled event needs to be handled, depends on initialization 94, and depends on watchdog 92. Components that depend on context switcher 100 are LED control 108, self test 102, single line display 106, SLC injection logic 110, serial 110 88, and executive loop 90. As used herein, the term “computer” includes any processor-based or microprocessor-based system including systems using microcontrollers, reduced instruction set circuits (RISC), application specific integrated circuits (ASICs), logic circuits, and any other circuit or processor capable of executing the functions described herein. The above examples are exemplary only, and are thus not intended to limit in any way the definition and/or meaning of the term “computer”. As used herein, the terms “software” and “firmware” are interchangeable, and include any computer program stored in memory for execution by a computer, including RAM memory, ROM memory, EPROM memory, EEPROM memory, and non-volatile RAM (NVRAM) memory. The above memory types are exemplary only, and are thus not limiting as to the types of memory usable for storage of a computer program. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims. |
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047626763 | abstract | A top nozzle adapter plate for use in a fuel assembly of a nuclear reactor includes an upper structural component and a lower functional component supported from the upper component. The fuel assembly has elongated structural members and fuel rods disposed in a predetermined array. The upper structural component of the adapter plate includes spaced and interconnected hubs and ligaments arranged to define substantial open areas for coolant flow therethrough while providing a rigid framework capable of transmitting lifting loads imposed by the fuel assembly. The hubs are connected to the elongated structural members of the fuel assembly. The lower functional component of the adapter plate includes a grid composed of a plurality of spaced and interleaved straps which cross one another to form intersections aligned with individual fuel rods in the array thereof for restraining movement of fuel rods upward from the fuel assembly while defining open channels through the grid for allowing passage of coolant flow therethrough. The grid also contains void areas through which extend the hubs of the upper component. The lower component also includes coolant flow directing means being operable to establish a predetermined desired pressure drop across the top nozzle of the fuel assembly. In one form, the coolant flow directing means is a plurality of tabs connected to the grid straps and extending outwardly therefrom. The tabs are adjustable into various desired positional relationships with respect to the grid channels for controlling coolant flow therethrough. In a modified form, the coolant flow directing means is in the form of a thin flat plate having holes of predetermined desired sizes and shapes formed therein. The plate extends along the interleaved straps of the grid with its holes generally aligned with the open channels of the grid. |
claims | 1. A method of performing an optical proximity correction (OPC) procedure, comprising:forming an integrated chip (IC) design comprising a graphical representation of an integrated chip, wherein the IC design has an original design layer comprising one or more original design shapes;generating an approximation design layer from the original design layer, wherein the approximation design layer is different than the original design layer; andperforming an optical proximity correction (OPC) procedure on the approximation design layer, by dissecting the approximation design layer into a plurality of separate edges and then selectively moving one or more of the separate edges in a direction perpendicular to the separate edge, to form an OPC'd layer that produces on-wafer shapes that correspond to the one or more original design shapes,wherein a computing device is used to form the IC design, to generate the approximation design layer, or to perform the OPC procedure on the approximation design layer. 2. The method of claim 1, wherein generating the approximation design layer, comprises:modifying the original design layer to replace an angled edge with a vertical edge having a 90° slope and a horizontal edge having a 0° slope. 3. The method of claim 1, further comprising:writing the OPC'd layer generated from the approximation design layer onto a photomask. 4. The method of claim 3, wherein the OPC'd layer written onto the photomask comprises a horizontal edge and a vertical edge at positions corresponding to a 45° edge having a substantially 45° slope within the original design layer. 5. The method of claim 1, wherein performing the OPC procedure comprises:separating edges of the approximation design layer into plurality of distinct edges that are contiguously connected; andselectively moving one or more of the plurality of distinct edges to adjust a shape of the approximation design layer in a manner that mitigates optical proximity effects. 6. The method of claim 5, wherein the one or more of the plurality of distinct edges are selectively moved in a non-perpendicular direction with respect to corresponding edges of the original design layer. 7. The method of claim 1, wherein the original design layer is a subset of the approximation design layer. 8. The method of claim 1,wherein the original design layer comprises mirror imaged 45° edges separated by an interconnecting vertical edge, andwherein the OPC procedure is configured to move the mirror imaged 45° edges and the interconnecting vertical edge in an inward direction or in an outward direction opposite the inward direction. 9. A method of performing an optical proximity correction (OPC) procedure, comprising:forming an integrated chip (IC) design comprising a graphical representation of an integrated chip having an original design layer comprising one or more original design shapes corresponding to structures that are to be formed on an integrated chip;generating an approximation design layer, from the original design layer, by replacing a 45° edge having a substantially 45° slope in the original design layer with a vertical edge having a 90° slope and a horizontal edge having a 0° slope in the approximation design layer;separating edges of the approximation design layer into a plurality of distinct edges that are contiguously connected; andselectively moving the distinct edges to adjust a shape of the approximation design layer to form an OPC'd layer that produces on-wafer shapes that correspond to the one or more original design shapes,wherein a computing device is used to form the IC design, to generate the approximation design layer, to separate edges of the approximation design layer, or to selectively move the distinct edges. 10. The method of claim 9, further comprising:writing the OPC'd layer generated from the approximation design layer onto a photomask. 11. The method of claim 10, wherein the OPC'd layer written onto the photomask comprises the horizontal edge and the vertical edge at positions corresponding to the 45° edge in the original design layer. 12. The method of claim 11, wherein the one or more of the plurality of distinct edges are selectively moved in a non-perpendicular direction with respect to corresponding edges of the original design layer. 13. The method of claim 11, wherein the original design layer is a subset of the approximation design layer. 14. The method of claim 9, further comprising:wherein the original design layer comprises mirror imaged 45° edges separated by an interconnecting vertical edge, andwherein the OPC procedure is configured to move the mirror imaged 45° edges and the interconnecting vertical edge in an inward direction or in an outward direction opposite the inward direction. 15. An electronic design automation (EDA) tool, comprising:a memory element comprising an electronic storage unit configured to store an integrated chip (IC) design comprising a graphical representation of an integrated chip, wherein the IC design has an original design layer comprising one or more original design shapes;an approximation design generation element comprising a first computing device configured to generate an approximation design layer from the original design layer, wherein the approximation design layer is different than the original design layer; andan OPC element comprising a second computing device configured to perform an optical proximity correction (OPC) procedure on the approximation design layer, la dissecting the approximation design layer into a plurality of separate edges and then selectively moving one or more of the separate edges in a direction perpendicular to the separate edge, to form an OPC'd layer that produces on-wafer shapes that correspond to the one or more original design shapes. 16. The EDA tool of claim 15, wherein generating the approximation design layer, comprises:modifying the original design layer to replace a 45° edge having a substantially 45° slope with a vertical edge having a 90° slope and a horizontal edge having a 0° slope. 17. The EDA tool of claim 15, further comprising:a mask writing tool configured to write the OPC'd layer generated from the approximation design layer onto a photomask. 18. The EDA tool of claim 15, wherein the OPC element is configured to:separate edges of the approximation design layer into a plurality of distinct edges that are contiguously connected; andselectively move one or more of the plurality of distinct edges to adjust a shape of the approximation design layer in a manner that mitigates optical proximity effects. 19. The EDA tool of claim 18, wherein the one or more of the plurality of distinct edges are selectively moved in a non-perpendicular direction with respect to corresponding edges of the original design layer. 20. The EDA tool of claim 15,wherein the EDA tool is configured to receive the integrated chip design from a design tool and to store the integrated chip design in the memory element; andwherein the EDA tool is configured to process the received integrated chip design to perform the OPC procedure. |
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claims | 1. A method for producing and extracting a radioisotope fission product, the method including:(a) providing a target including solid phase uranium oxide;(b) irradiating, by a nuclear reactor, the target with neutrons so as to convert a first portion of the solid phase uranium oxide into oxides of radioisotope fission products;(c) gaseously extracting at least one of the oxides from the irradiated target of step (b), said extracting comprising halogenating the at least one of the oxides using an extraction gas including Cl2 gas so as to produce at least one gas phase oxidized and halogenated radioisotope fission product, wherein the at least one gas phase oxidized and halogenated radioisotope fission product includes molybdenum-99, and so as to produce a first residual target including a remainder of the irradiated target of step (b) less the at least one of the oxides extracted;(d) irradiating, by the nuclear reactor, the first residual target of step (c) with neutrons so as to convert a second portion of the solid phase uranium oxide into oxides of the radioisotope fission products; and(e) gaseously extracting the at least one of the oxides from the irradiated first residual target of step (d), said extracting comprising halogenating the at least one of the oxides using the extraction gas so as to additionally produce at least one gas phase oxidized and halogenated radioisotope fission product. 2. The method of claim 1, wherein the solid phase uranium oxide provided in step (a) has a U-235 enrichment above about 20%. 3. The method of claim 1, wherein the solid phase uranium oxide provided in step (a) has a U-235 enrichment less than about 20%. 4. The method of claim 1, wherein the solid phase uranium oxide provided in step (a) includes UO3 or U3O8. 5. The method of claim 1, wherein step (b) comprises irradiating the target in the nuclear reactor. 6. The method of claim 5, wherein step (c) comprises removing the target is from within the reactor for the extracting of step (c). 7. The method of claim 1, wherein step (b) comprises substituting the target for a nuclear reactor fuel element. 8. The method of claim 1, wherein the target provided in step (a) includes a primary containment. 9. The method of claim 1, wherein the target provided in step (a) comprises:a length of cladding having upper and lower ends and defining an interior space;upper and lower endcaps sealed to the upper and lower ends; andat least one gas port in fluid communication with the interior space; andwherein the uranium oxide is disposed within the interior space and has a form selected from the group consisting of powder, granules, or porous annular pellets. 10. The method of claim 9, wherein the target provided in step (a) further comprises a barrier layer disposed on an interior surface of the cladding. 11. The method of claim 9, wherein the cladding of the target provided in step (a) includes silicon carbide or quartz. 12. The method of claim 1, wherein the extracting of step (c) comprises converting the molybdenum-99 to molybdenum oxychloride (MoO2Cl2), a gaseous species. 13. The method of claim 1, the extraction gas of step (c) further including an oxygen-containing species selected from the group consisting of Ox, H2O, NOx, COx and ClOx, where x can take on any chemically permissible value. 14. The method of claim 1, wherein the extracting of step (c) comprises extracting a plurality of gaseous species. 15. The method of claim 1, wherein step (c) further comprises:using a gas inflow system coupled to the target to introduce the extraction gas into the target;heating the target to promote halogenating the at least one of the oxides using the extraction gas;using a gas outflow system coupled to the target to transfer the at least one gas phase oxidized and halogenated radioisotope fission product from the target; andcollecting the transferred at least one gas phase oxidized and halogenated radioisotope fission product in a recovery chamber. 16. The method of claim 15, wherein the transferred at least one gas phase oxidized and halogenated radioisotope fission product of step (c) includes molybdenum oxychloride (MoO2Cl2). 17. The method of claim 15, wherein the target is heated during step (c) to a temperature in a range from about 200 C to about 1500 C. 18. The method of claim 15, wherein during step (c) the at least one gas phase oxidized and halogenated radioisotope fission product is transferred out of the target to the recovery chamber at a temperature sufficiently high to inhibit solidification of the at least one gas phase oxidized and halogenated radioisotope fission product during the transfer. 19. The method of claim 15, wherein during step (c), introducing the extraction gas into the target comprises continuously flowing the extraction gas into the target during the steps of converting, transferring, and collecting. 20. The method of claim 15, wherein during step (c), introducing the extraction gas into the target comprises initial introduction of the extraction gas, a period of no flow of the extraction gas during formation of the at least one gas phase oxidized and halogenated radioisotope fission product, and a resumed flow of the extraction gas to extract and transfer the at least one gas phase oxidized and halogenated radioisotope fission product. 21. The method of claim 15, step (c) further comprising, after transferring the at least one gas phase oxidized and halogenated radioisotope fission product out of the target, filling the target with a fill gas, and returning the target to the nuclear reactor for the irradiating of step (d). 22. The method of claim 1, wherein step (b) comprises irradiating the target adjacent to the nuclear reactor. 23. The method of claim 22, wherein step (c) comprises the target remaining adjacent to the nuclear reactor during the extracting of step (c). 24. The method of claim 22, wherein step (c) comprises removing the target from adjacent to the nuclear reactor for the extracting of step (c). |
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claims | 1. A neutron generator comprising:a housing defining an ion source chamber;an electron nano-emitter arrangement comprising multiple field electron emitters located in the ion source chamber;an electrode arrangement comprising an anode separated from the electron nano-emitter arrangement by a separation space, wherein the separation space holds an ionizable gas; anda permanent magnet that provides a magnetic field in misalignment with an electric field generated by the electrode arrangement. 2. The neutron generator of claim 1, wherein the housing has a longitudinal axis, the electron nano-emitter arrangement being oriented to emit electrons along a path parallel to the longitudinal axis, wherein the anode is not in the path parallel to the longitudinal axis. 3. The neutron generator of claim 2, wherein the electron nano-emitter arrangement comprises a substrate providing a substantially circular substrate on which the field electron emitters are supported, the substrate being co-axial with the longitudinal axis of the housing and extending transversely thereto. 4. The neutron generator of claim 2, wherein the magnetic field is aligned perpendicular to the longitudinal axis. 5. The neutron generator of claim 2, wherein the permanent magnet is coaxial with the longitudinal axis of the ion source chamber. 6. The neutron generator of claim 1, wherein the permanent magnet and the multiple field electron emitters are located at a common longitudinal end of the housing. 7. The neutron generator of claim 2, wherein the anode is substantially annular and co-axial with the longitudinal axis of the ion source chamber, the anode being axially spaced from the multiple field electron emitters. 8. The neutron generator of claim 3, wherein an inner radius of the anode is greater than a radially outer extremity of the substantially circular substrate of the multiple field electron emitters. 9. The neutron generator of claim 1, further comprising a control arrangement to apply an ion source voltage pulse between the electron nano-emitter arrangement and the anode to generate the electric field, the electron nano-emitter arrangement serving at least in part as an ion source cathode. 10. The neutron generator of claim 9, wherein the ion source voltage pulse has an amplitude of 500 V or less. 11. The neutron generator of claim 9, wherein a turn-on/turn-off delay for an ion source comprising the electron nano-emitter arrangement and the anode is smaller than 1 μs. 12. The neutron generator of claim 1, wherein the field electron emitters comprise multiple nanotips, each nanotip comprising an elongated filament which is substantially aligned with a longitudinal axis of the housing. 13. The neutron generator of claim 1, further comprising: a target structure holding target particles, wherein the target particles comprise isotopes. 14. The neutron generator of claim 13, wherein the target particles comprise at least one of deuterium or tritium. 15. A neutron generator comprising:a housing defining an ion source chamber;an electron nano-emitter arrangement comprising multiple field electron emitters located in the ion source chamber;an electrode arrangement comprising an anode separated from the electron nano-emitter arrangement by a separation space by a separation space, wherein the separation space holds an ionizable gas, and wherein the anode is substantially annular and co-axial with a longitudinal axis of the ion source chamber; anda permanent magnet that provides a magnetic field in misalignment with an electric field generated by the electrode arrangement. 16. The neutron generator of claim 15, wherein the permanent magnet is arranged with respect to the electrode arrangement to provide the magnetic field substantially perpendicular to the electric field in at least one part of the ion source chamber. 17. The neutron generator of claim 15, wherein the permanent magnet is arranged to provide the magnetic field aligned perpendicular to the longitudinal axis of the ion source chamber. 18. The neutron generator of claim 15, further comprising a target structure positioned longitudinally opposed to the electron nano-emitter arrangement, wherein the target structure comprises one or more isotopes, and wherein the separation space is between the target structure and the electron nano-emitter arrangement. 19. The neutron generator of claim 18, wherein the isotopes comprise at least one of tritium and deuterium. 20. A neutron generator comprising:a housing defining an ion source chamber;an electron nano-emitter arrangement comprising multiple field electron emitters located in the ion source chamber;an electrode arrangement comprising an anode separated from the electron nano-emitter arrangement by a separation space, wherein the separation space holds an ionizable gas; anda permanent magnet that provides a magnetic field in misalignment with an electric field generated by the electrode arrangement, wherein the permanent magnet and the multiple field electron emitters are located at a common longitudinal end of the housing. |
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abstract | The fuel element of the present invention includes a cladding tube and a metal fuel contained in the cladding tube, in which a gas plenum region is formed above the metal fuel and inside the cladding tube and has a small-diameter portion in the gas plenum region. Further, the fuel assembly of the present invention includes the fuel element of the present invention and a wrapper tube surrounding the fuel element, in which a coolant material passage is formed between the fuel element and the fuel element. Further, the core of the present invention includes an inner core fuel region loaded with the fuel assembly according to the present invention, and an outer core fuel region loaded with the fuel assembly of the present invention. |
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summary | ||
050911406 | claims | 1. A method of replacing a damaged heater nozzle in a nuclear reactor coolant system pressurizer wherein an electric heater extends into the pressurizer through the heater nozzle and bore in the wall of the pressurizer, comprising: a. removing the electric heater; b. removing the damaged heater nozzle; c. enlarging the bore in the wall of the pressurizer; d. installing an outer sleeve in the enlarged bore by welding it to the inner and outer surfaces of the pressurizer; e. installing an inner sleeve inside the outer sleeve so as to extend into the pressurizer beyond the upper end of said outer sleeve and welding said inner sleeve to the lower end of said outer sleeve; and f. installing an electric heater so as to extend through said inner sleeve into said pressurizer and welding said electric heater to the lower end of said inner sleeve. 2. The method of claim 1, wherein said outer sleeve is substantially flush with the interior of said pressurizer. |
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