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abstract | An ion implanter includes an ion source for generating an ion beam moving along a beam line and a vacuum or implantation chamber wherein a workpiece, such as a silicon wafer is positioned to intersect the ion beam for ion implantation of a surface of the workpiece by the ion beam. A liner has an interior facing surface that bounds at least a portion of the evacuated interior region and that comprises grooves spaced across the surface of the liner to capture contaminants generated within the interior region during operation of the ion implanter. |
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058898310 | claims | 1. A containment of a nuclear power station, comprising: a device for igniting hydrogen contained in a hydrogen/air mixture, said device including a central electrode for lightning flash generation and a high-voltage source connected to said central electrode for generating a high voltage greater than a disruptive discharge voltage of air. 2. The containment according to claim 1, wherein said high-voltage source generates a high voltage with a frequency of more than 1 kHz. 3. The containment according to claim 1, including a number of counter-electrodes disposed in an interior of the containment for conducting lightning flashes. |
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053316746 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Before explaining the invention in detail, it is to be understood that the invention is not limited in its application to the detail of construction and arrangement of parts illustrated in the drawings since the invention is capable of other embodiments and of being practiced or carried out in various ways. It is also to be understood that the phraseology or terminology employed is for the purpose of description only and not of limitation. In order to enable the reader to better understand the following description of the invention, the following list provides typical water and system component elevations for a water cooled reactor and steam generator shown in FIG. 1. Values not pertinent to the explanation of the present invention are denoted not applicable, "N/A". ______________________________________ drawing feet Component designator elevation ______________________________________ refuel deck 16 2047.5 N/A mid loop level 70 2014.5 residual heat removal system RHR 90 2013.7 suction line top of fuel 94 2010.3 RHR discharge 18 1971.8 N/A to reactor (mid loop level) 70 2014.5 nozzle dam 20 2017 RHR loop 74 -- refueling level 80 2044.6 pressurizer 10 N/A steam generator 46 -- reactor 96 -- reactor coolant pump, RCP 38 N/A RHR pump 92 -- ______________________________________ Referring to FIGS. 1 and 2, nozzle dam 20 is installed over nozzle 26 of hot leg 30 of the hot side off partition 34 of bowl 40 of steam generator 46. Bowl drain 50 includes passage 54 which connects nozzle 26 with bowl 40, bypassing the hermetic seal 56 that is provided by the nozzle dam when in intimate contact with the nozzle wall 60. Drain 50 is sealed by screw-in plug 66. After the reactor coolant system "RCS" inventory water general level is lowered to mid loop level 70 of the residual heat removal system "RHR" loop 74 and of the hot and cold primary coolant inlet 30 and outlet 32 legs respectively of the steam generator bowl, nozzle dams are installed in the hot leg and cold leg nozzles, and plugs are installed in the individual bowl drains. After all nozzle dams are installed, the RCS inventory level is raised to refueling level 80 which causes a pressure head of about 27 feet of water (11.7 psig) directed upward against nozzle dam 20, and a pressure head of 43.5 feet (18.9 psig) at the lowest elevation of the cold leg, assuming that lowest elevation to be about 14 feet below mid loop in most plants. Referring additionally to FIG. 3, the nozzle dam inflatable seals 82 are each maintained at approximately 65 psig. Annulus 84 between the inflatable seals is maintained at 5 psig and monitored in order to learn of leakage in the seal system. Slow increase has been widely observed in annulus pressure. This may indicate that a very small amount of air can escape from the inflatable seals either by osmosis or from minute leaks. It is now believed by the present inventors that this air can also leak to the nozzle side of the seal. Pressurized seals presently seem to be the most efficient and safest way to seal a nuclear reactor steam generator nozzle. It is also now believed by the inventors that prolonged installation of a nozzle dam having an inflatable seal could allow enough osmotic or small leakage of air from the 65 psig source to displace the 11.7 to 18.9 psig water occupying the volume of the cold leg below the cold leg nozzle elevation. After having displaced the water, continuing air bubbles out toward RCS inventory storage. The nozzle dams remain in place while maintenance is completed on the reactor system, sometimes for as long as 30 days, then the water level is dropped from refueling level 80 to mid loop 70 in order to remove the nozzle dams. This reduces the pressure on the trapped column of air in the cold leg to a head of 14 ft. The trapped air exerts a pressure of approximately 6.1 psig totaling 7,662 pounds against the underside of the 40" diameter nozzle dam. Under this pressure, upward bounding can occur, driven by the trapped air as soon as a movable portion of a nozzle dam is unbolted. In a nozzle dam system relying solely upon inflatable seals, rather a combination passive and active inflatable system as in the BUSI Nozzle Dam, the trapped air is released around the dam when the seals are deflated, and bounding does not occur. Nevertheless loss of RCS inventory level can still occur. For steam generators with individual drain lines, bounding can be reduced by removing the drain plug after the reactor coolant system water general level is moved to mid loop level, and waiting for some period of time before removing the nozzle dam. Nevertheless loss of RCS inventory level can still occur. In steam generators with a common drain instead of individual drain lines, there is no access to the nozzle region just below the nozzle dam. The cause for loss of RCS inventory level is considered to be as follows. The volume of the trapped air is conservatively calculated to be 15.75 inches radius squared (or 248.06 square inches), times 3.14 (Pi), divided by 144, times 16.5 feet vertical height=89.25 cu. ft. or 669 gallons. When the cold leg nozzle dam is removed, the trapped air escapes and water rushes upward to equalize the reactor coolant system water general level, the RCS inventory could suddenly decrease by 669 gallons. Assuming 1191 gallons per vertical foot of RCS inventory, this event could result in a sudden decrease in RCS inventory level by about 7.2 inches. The residual heat removal system RHR suction line 90 is only 9.6 inches below mid loop 70. Taking in the above conservative calculation plus bends in cold legs, increase in volume near nozzles and water sloshing effects, it is conceivable that a single occurrence of this event can cause the localized inventory level close to the RHR suction line to fall, causing a risk of vortexing/cavitation failure of RHR pump, and subsequent loss of effectiveness of the RHR. Furthermore, unless RCS inventory is recovered after removal of the first cold leg dam of a system, further substantial risk exists when a second cold leg dam is removed, assuming that the second cold leg nozzle dam experiences a similar occurrence. If four cold legs in a system experience the same occurrence without water level recovery, the water level would drop to about 21 inches from the top 94 of the fuel in reactor 96. In order to avoid the above problem associated with use of nozzle dams with inflatable seals, it is advisable to bleed the trapped air from the cold leg before the RCS inventory drain down process is completed, so that the water level in the cold leg seeks the RCS inventory water general level. This process step can be accomplished preferably by passing the air from the nozzle leg into the bowl by way of a valve and passage through the nozzle dam. Another way is by passing the air from the nozzle leg into the bowl by way of a valve connected to passage 54 in place of the presently used plug. Referring to FIG. 4, valve assembly 100 passes through passage 108 of nozzle dam 20 body wall 102. If the nozzle dam is a BUSI Nozzle Dam, the valve is preferably mounted in the center section of the dam's three sections. Flared end 106 of fitting 104 sealingly engages rubber diaphragm 110 which spans the three sections. At the other end of fitting 104 is pipe means such as 1/4 inch reinforced hose 112 securely attached to the fitting. Adaptor 114 permits connection of hose 112 before the water level is to be lowered. Connector section 116 seals fitting 104 until hose 112 is attached for use of the fitting. Control valve 118, connected to passage 108 by way of hose 112 and fitting 104 is preferably located outside the generator bowl so that it can be easily controlled by an operator. Water-stop gas-conducting valve 130 releases the trapped air which flows via passage 108 from the nozzle in the region immediately below the nozzle dam, remaining open until it encounters water which it blocks. This valve design may be taken from ones presently used to automatically bleed air from water circulator systems. A valve assembly 118 similar to valve assembly 100 may be tapped into passage 54 by using a suitably modified fitting as shown in FIG. 5. In operation, valve 18 is turned on or opened shortly before or during lowering of the water level toward the mid loop level, and air is permitted to flow from the pipe until water reaches the valve. It is then known that the compressed air is removed, and valve 118 is turned off or closed. Automatic valve 130 can perform the same function. The hose is preferably positioned for directing the air out of the bowl. It may be for example passed out of the bowl through a manway, or connected to a bowl common drain. The hose may also be extended to a height that is higher than the level of the RCS water general level at which the air bleed is taking place. Visual or sensor indication may then be had from the raised hose to determine when the hydrostatic level of the water under the dam is equal with that of the RCS water general level. When the trapped air is removed, and the water general level is brought to mid-loop below the height of the nozzle dam, the dam may then be removed. Although the invention has been described in terms of specific preferred embodiments, it will be obvious to one skilled in the art that various modifications and substitutions are contemplated by the invention disclosed herein and that all such modifications and substitutions are included within the scope of the invention as defined in the appended claims. |
abstract | The present invention relates to a degreasing composition, to a liquid, to a gel and to a degreasing foam which comprise said composition. |
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abstract | A sheathed, annular metal fuel system is described. A metal fuel pin system is described that includes an annular metal nuclear fuel alloy. A sheath may surround the metal nuclear fuel alloy, and a cladding may surround the sheath. A gas plenum may also be present. Mold arrangements and methods of fabrication of the sheathed, annular metal fuel are also described. |
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053176073 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to a crack removal tool. More specifically, this invention relates to a crack removal tool which utilize EDM techniques and which are suitable for removing cracks which appear within conduits, nozzles and the like, such as are found in nuclear reactor related structures. 2. Description of the Related Art During inspection of certain types of nozzle penetrations in pressurized water nuclear reactor heads, cracks may be discovered in weld areas of one or more nozzles. Access to these defects is severely hindered by the presence of thermal sleeves which are welded in place in nozzles and which prevent access with tooling. Accordingly, it is extremely difficult to effect repairs. SUMMARY OF THE INVENTION It is an object of the present invention to provide cutting and removal tools or heads which simplify the removal and repair of cracks which tend to develop in nozzle penetrations and similar types of conduiting which are used in nuclear reactors and the like type of environments. In brief, the above object is achieved through the use of EDM (electro discharge machining) heads which were developed so as to permit the cutting and partial removal of a thermal sleeve so as to expose a cracked area to permit defect blending within the head penetration In brief, the above object is achieved through the use of a tool system which features an EDM crack removal head which can be positioned within the nozzle or thermal sleeve and then oscillated in vertical and horizontal directions to enable precise material removal, leaving a surface free from cracks and in a condition for rewelding if required. In the event that a crack is found in a surface which is covered by a thermal sleeve, a small specialized EDM cutting head which can be delivered from an under side of the reactor head, and which has the ability to pass through the ID of the thermal sleeve, is used to cut the thermal sleeve and allow a portion of the sleeve to be removed. The crack can then be readily removed using the crack removal type of head. The use of the tooling system according the present invention enables repair of cracked nozzles in PWR heads from the underside, thus eliminating a costly need to remove a control rod drive mechanism and associated hardware from a reactor. More specifically, a first aspect of the present invention resides in a tool for use with conduiting, comprising: an elongated shaft structure; first and second support members disposed on the shaft structure which engage an inner wall of a conduit when actuated; cutting head means, supported on the shaft structure between the first and second support members, for performing a cutting operation on an inner wall of the conduit, the cutting head means including a movable cutting electrode and first servo means for selectively moving the electrode member laterally outwardly with respect the shaft structure; and second servo means, operatively connected with the cutting head mean for selectively displacing the cutting electrode in at least one of first and second rotational directions. A second aspect of the present invention resides in a crack removal tool comprising: an elongate shaft structure; first and second inflatable members disposed on the shaft structure; a crack sensor supported on the shaft structure at a location between the first and second inflatable members; a crack removal head supported on the shaft between the first and second inflatable members at a predetermined distance from the crack sensor, the crack removal head including a removal electrode and electrode servo means for selectively displacing the removal electrode radially with respect to an axis of the shaft structure; and head servo means for selectively displacing the crack sensor and the crack removal head with respect to the first and second inflatable members and for causing the crack removal head to undergo rotation about the axis of the shaft structure in at least one of first and second rotational directions. A third aspect of the present invention resides in a sleeve cutting tool comprising: an elongate shaft structure; first and second inflatable members disposed on the shaft structure; a cutting head supported on the shaft between the first and second inflatable members, the cutting head including a cutting electrode and cutting electrode servo means for selectively displacing the cutting electrode radially with respect to an axis of the shaft structure; and cutting head servo means for selectively displacing the cutting head in at least one of first and second rotational directions. A fourth aspect of the present invention resides in a crack removal system comprising: a crack removal tool which comprises: a first elongated shaft structure; first and second selectively actuatable support members disposed on the shaft structure; a crack sensor supported on the shaft structure at a location between the first and second support members; a crack removal head supported on the shaft between the first and second support members at a predetermined distance from the crack sensor, the crack removal head including a crack removal electrode and a first servo means for selectively displacing the crack removal electrode radially with respect to an axis of the shaft structure; and second servo means for selectively axially displacing the crack sensor and the crack removal head with respect to one of the first and second support members and for causing the crack removal head to undergo rotation about the axis of the shaft structure in at least one of first and second rotational directions. A fifth aspect of the present invention resides in a crack removal system further comprising: a sleeve cutting tool which comprises: a second elongate shaft structure; third and fourth selectively actuatable support members disposed on the second shaft structure; a cutting head supported on the shaft between the third and fourth support members, the cutting head including a cutting electrode and third servo means for selectively displacing the cutting electrode radially outwardly with respect to an axis of the second shaft structure; and fourth servo means for selectively causing the cutting head to undergo rotation about the axis of the shaft structure in at least one of first and second rotational directions. A further aspect of the present invention is presented in a method of repairing cracks in a nozzle in which a sleeve is fixedly disposed, comprising the steps of: inserting sleeve cutting tool into the sleeve; activating support means which engages the inner wall of the sleeve and which maintains the sleeve cutting tool in a predetermined position within the sleeve; cutting through the sleeve using the sleeve cutting tool in a manner which allows a portion of the sleeve to be removed; removing the portion of the sleeve; inserting a crack removing tool into the portion of the nozzle wherein the sleeve has been removed; activating support means on the crack removing tool to support the crack removing tool in the nozzle; using a crack detecting sensor to locate the position of a crack in the nozzle; moving the crack detecting sensor away from the location at which the crack is detected and moving a crack removal head into position opposite the location whereat the crack was detected; and removing portion of the inner wall of the nozzle using the crack removal head. |
046438700 | summary | 2. Field of the Invention This invention relates to nuclear reactors, and particularly, to how nuclear reactor containment structures may dissipate heat following an extremely low probability meltdown and subsequent breach in the reactor vessel. 3. Background Discussion Because radioactive materials are contained in a nuclear reactor, great caution must be taken to prevent the escape of such materials to the environment. One type of nuclear reactor is the liquid metal fast-breeder reactor which employs a core immersed in liquid sodium coolant. If all heat removal capacity were lost and the temperature within the reactor should exceed the melting point of the core, the core would disintegrate and core materials could reach the bottom of the reactor, where the debris layer heat generation rate could be sufficiently high to melt the walls of the reactor vessel and guard vessel. If this would occur both sodium and fragmented, radioactive core debris would escape from the reactor vessel. The reactor containment must be designed to retain such radioactive materials which might penetrate the reactor vessel, and prevent their entry into the environment where they can endanger public health & safety. The present invetion provides a safe containment structure which has the advantages of a system for cooling the containment structure and retained core debris to provide protection in the extremely unlikely event of a breach of the reactor vessel, so that failure of the containment by the core debris interactions is avoided. BRIEF DESCRIPTION OF THE INVENTION The nuclear reactor of this invention includes a reactor vessel disposed in a thick-walled metal cavity. This cavity is preferably lodged partially or completely below the surface of the earth. Preferably it is tied to adjacent structural concrete. Preferably, the reactor vessel is seated in a guard vessel and both of these are seated within the surrounding metal cavity. There is a thick metal basemat beneath the reactor vessel at the bottom of the cavity. The cavity wall, at the bottom abutting the basemat, is welded or otherwise integral with the metal basemat. Disposed in the zone below the basemat and adjacent to it, are means for feeding water into this region. The zone immediately underneath the base plate is composed of porous media such as sand and gravel. If there is a core meltdown and subsequent breach in the reactor vessel, this zone below the basemat will eventually be heated above the boiling point of water. Water fed into the heated zone will be converted into steam. Its latent heat of vaporization provides cooling of both the basemat and contained core debris. Means are provided for venting the steam to the atmosphere as it is formed. The metal base plate preferably extends beyond the perimeter of the cavity wall and it is supported on metal pilings which preferably extend downwardly and outwardly into the earth. The metal pilings serve to conduct heat away from the reactor into water-saturated porous media and earth beneath. At the same time they support the reactor and prevent it from sinking into the earth. |
claims | 1. A fault monitoring method for a robot or other work machine wherein output from a servo motor is transmitted via a reducer or other power transmission mechanism and a load is driven, the method comprising the steps of:acquiring first torque data generated from the servo motor, in units each of which starts when the motor starts operating and ends when the motor stops operating;selecting a maximum first torque fluctuation range designated by a difference between a maximum torque and a minimum torque for each unit obtained from the acquired first torque data;collecting maximum fluctuation ranges of the first torque for a plurality of cycles for obtaining a first average value of the maximum fluctuation ranges of the first torque;selecting a fluctuation range control value by multiplying the first average value by a factor greater than 1.0;acquiring second torque data generated from the servo motor, in units each of which starts when the motor starts operating and ends when the motor stops operating, after the fluctuation range control value has been selected;selecting a maximum second torque fluctuation range designated by a difference between a maximum torque and a minimum torque among the units obtained from the acquired second torque data;collecting second maximum fluctuation ranges for a plurality of cycles for obtaining a second average value of the second maximum fluctuation ranges;making a comparison to determine whether the second average value exceeds the fluctuation range control value; anddetermining, by a fault monitoring device, that a fault has occurred when the second average value exceeds the control value for the fluctuation range in the comparison. 2. The method of claim 1, wherein a plurality of servo motors is provided, a fluctuation range control value is set for each of the servo motors, and the servo motors are centrally controlled by a single checking unit. 3. The method of claim 2, wherein the acquired torque data is converted to consolidated data by a data converter and then sent to the checking unit. 4. The method of claim 1, wherein the first torque data and second torque data are acquired from a motor driver or controller for controlling the servo motor. 5. The method of claim 1, wherein a warning signal is generated from an alerting unit when a fault is determined in the step for determining that a fault has occurred. |
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052727335 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 represents a control rod driving hydraulic system relating to a first embodiment. The embodiment comprises replacing the driving water filters 6 installed on an outlet side of the driving water pump 4 shown in FIG. 8 as a prior art with hollow fiber filter units 28. Referring to FIG. 1, during operation of a nuclear power plant, water coming from the condensate demineralizer 3 passes the suction filter 5, and after being pressurized by the driving water pump 4, it passes the hollow fiber filter unit 28, and is partly supplied to the control rod driving mechanism 1 by way of the flow control valve 8 and the pressure control valve 10. Here, foreign materials such as cladding and others which are mixed into the driving water are removed first from passing the suction filter 5. However, a minute insoluble solid material of about 25 .mu.m or below in size cannot be removed through the suction filter 5 and thus comes into the driving water pump 4, and is discharged to the hollow fiber filter unit 28, and the solid material is removed almost perfectly. Accordingly, the driving water which comes out of the hollow fiber filter unit 28 becomes pure and free from foreign materials therein. Thus, an abnormity of extra ordinary operation of the control rod driving mechanism 1 due to biting of foreign materials in the driving water is prevented from arising to maintain a stable driving performance. In the case of the prior art exemplified in FIG. 8, when equipment is disassembled for plant construction and periodical survey or inspection and a water source is changed from the condensate demineralizer 3 to the condensate storage tank 18, a work for withdrawing the air must be carried out with great care, however, since the hollow fiber filter unit 28 is installed on an outlet (discharge) side of the driving water pump 4 of the control rod driving hydraulic supply system in the embodiment, the air is collected through the hollow fiber filter unit 28. Thus, the aforementioned vent ,work can sharply be cut down and simplified as well, and therefore a time consuming construction process and a periodical survey process can be reduced, thereby realizing an economical management. FIG. 2 represents a second embodiment of the control rod driving hydraulic system relating to the present invention, and this embodiment comprises replacing the suction filters 5 provided on an inlet side of the driving water pump 4 in the case of prior art exemplified in FIG. 8 with the hollow fiber filter units 28. The remaining construction is not different from that of FIG. 8, Therefore a further description will be omitted here. A driving water from the condensate demineralizer 3 passes the hollow fiber filter unit 28 before entering the driving water pump 4, foreign materials being removed therethrough. Since the driving water having the foreign materials removed as above comes into the driving water pump 4 in this case, a performance of each equipment can be maintained as preventing the biting of the foreign materials to the driving water pump 4. The hollow fiber filter unit 28 collects the air as in the case of control rod driving device given in the first embodiment, therefore the air generated on an upstream side of the hollow fiber filter unit 28 is prevented from flowing into the piping, the control rod driving mechanism 1 and the water pressure control unit 2, and the vent work for withdrawing the air can sharply be cut to realize an economical management. In the above embodiments, the two driving water filters 6 and the two suction filters 5 are substituted with two hollow fiber filter units 28, respectively, only ones of them 6 and 5 may be substituted with the filter unit 28. FIG. 3 represents a control rod driving hydraulic system relating to a third embodiment according to the present invention, in which like reference numerals are added to elements or unit corresponding to those of the first and second embodiments and in which one of the driving water filters 6 in FIG. 8 installed on an outlet side of the driving water pump 4 is substituted with a hollow fiber filter 28. The other one of the filters 6 is of a conventional knotting-wire type filter. As shown in FIG. 4, a hollow fiber filter unit 28 is provided with a sealing casing 28a which is divided into upper and lower sections by a partition plate 28b provided with a number of perforations. Filter elements 28c each is composed of a number of hollow fibers each passing the perforation of the partition plate 28b. The driving water fed into a lower filtering chamber 28d of the sealing casing 28a is filtered by the filament elements 28c, and thereafter, the driving water is drained to an outlet unit 28h via an upper header chamber 28e and a drain pipe 28g provided with an open/close valve 28f. An inlet pipe 29a is connected to the driving water filter unit 28 so as to feed the driving water from the inlet unit 29, and a flow meter 29b and an open/close valve 29c are provided for the inlet pipe 29a. A differential monometer 29d is provided for a duct connecting the inlet pipe 29a and the drain pipe 28q. In this embodiment, water coming from the condensate storage tank 18 passes the suction filter 5 and after being pressurized by the driving water pump 4, it passes the hollow fiber filter unit 28 via the inlet pipe 29a and then, through the drain pipe 28g, is partly supplied to the control rod driving mechanism 1 by way of the flow control valve 8 and the pressure control valve 10. Here, foreign materials such as cladding and others which are mixed into the driving water is removed first from passing the suction filter 5. However, a minute insoluble solid material of about 25 .mu.m or below in size cannot be removed through the suction filter 5 and thus comes into the driving water pump 4, and is discharged to the hollow fiber filter unit 28 from the pump 4. A cladding of about 0.1 .mu.m or so is picked up through the hollow fiber filter unit 28 and the solid material is removed almost perfectly. Accordingly, the driving water which comes out of the hollow fiber filter unit 28 becomes pure free from foreign materials therein. Thus, an abnormity of extra ordinary operation of the control rod driving mechanism 1 due to the biting of the foreign materials in the driving water is prevented from arising to maintain a stable driving performance. The purge water for the mechanical seal of the reactor coolant recirculation pump 30 is branched from the control rod driving water system, so that when the control rod driving water is made clean, the purge water for the mechanical seal of the recirculation pump 30 can be also cleaned simultaneously. Accordingly, the biting of the foreign materials to the mechanical seal can be substantially prevented, thus reducing the leakage of water. In the case of the prior art exemplified in FIG. 8, when equipment is dissembled for plant construction and periodical survey or inspection and a water source is changed from the condensate demineralizer 3 to the condensate storage tank 18, a work for withdrawing the air must be carried out with great care, however, since the hollow fiber filter unit 28 is installed on an outlet (discharge) side of the driving water pump 4 of the control rod driving hydraulic supply system in the embodiment, the air is collected through the hollow fiber filter unit 28. Thus, the aforementioned vent work can sharply be cut down and simplified as well, loads of a construction process and a periodical survey or inspection process can be reduced, thereby realizing an economical management. The backwash regeneration equipment for the hollow fiber filter unit 28 will be described hereunder with reference to FIG. 4. Referring to FIG. 4, the backwash regeneration equipment 31 is provided with an air supply tube 33 led from an air inlet 32, and the air supply tube 33 is branched into two branch tubes on the way thereof, one being communicated with the header chamber 28e of the sealing casing 28 as an air purge duct 34 and the other being communicated with the filtering chamber 28d as a bubble supply duct, 35. The filtering chamber 28d is provided with a perforated bubble generation member 36, composed of a perforated tube, connected to the bubble duct 35. The air supply tube 33 is equipped with an open/close valve 37 and an air filter 38, and the air purge duct 34 and the bubble supply duct 35 are also equipped with open/close valves 39 and 40, respectively. A plurality, three, for example, of drain pipes 41, 42 and 43 are connected to the filtering chamber 28d of the hollow fiber filter unit 28 to portions of different levels, and these drain pipes 41, 42 and 43 are connected to a radioactive waste disposal unit 48 through open/close valves 44, 45, 46 and 47. In the hollow fiber filter unit 28 of the characters described above, during the water supply through the condensate storage tank 18, the quantity of the cladding caught by the fiber filament element 28c is gradually increased. This fact of increasing of the quantity of the cladding caught can be confirmed by the supply water quantity measurement by the flowmeter 29b and the pressure difference measurement by the differential manometer 29d between the inlet pipe 29a and the drain pipe 28g. When these measurement values reach predetermined reference values, it is discriminated that the washing of the hollow fiber filter unit 28 should be made. At the washing time of the hollow fiber filter unit 28, the driving water supply pump 4 first stops and the open/close valves 29c and 28f of the inlet pipe 29a and the drain pipe 28g are closed. In the next stage, the open/close valves 44 and 47, which are disposed for the drain pipes 41 and 43 disposed to high level positions of the hollow fiber filter unit 28, are made open thereby to drain the water in the unit 28 so that the water level therein is lowered below the partition plate 28b. In this operation, the open/close valves 37 and 39 of the air supply pipe 33 and the air purge duct 34 are made open and the pressurized air is supplied from the air inlet 32 to an upper portion of the sealing casing 28a of the hollow fiber filter unit 28, thus draining smoothly. The supply of the pressurized air makes swollen the filter elements 28c. Under this condition, the open/close valve 40 of the bubble supply duct 35 made open and the bubbling operation is performed in the filtering chamber 28d by the air supply to the bubble generation member 36. According to this bubbling, the cladding adhering on the outer surface of the filter elements 28c is removed, thus performing the backwash regeneration treatment. After the removal of the cladding by the bubbling for a predetermined time, the open/close valves 45 and 46 of the drain tube 42 are opened and the water in the filtering chamber 28d is drained into the radioactive waste disposal equipment 48. Thereafter, the valves opened for the backwash operation are closed and the valves such as the valve 29c opened during the backwash operation are then closed. The filtering operation of the driving water is then again carried out. According to this embodiment, the frequencies of the exchanging of the filter elements 28c of the hollow fiber filter unit 28 can be significantly reduced by this backwash regeneration process and the quantity of the secondary radioactive waste can be hence reduced. In addition, since it is possible to regenerate the filter elements 28c, it is not necessary to limit the water source to the condensate subjected to the demineralizing process carried out by the condensate demineralizing unit. In the meantime, according to this embodiment, one conventional knotting-wire type filter unit 6 and one hollow fiber filter unit 28 are arranged in parallel for two control rod driving water filters. Since it is necessary for the hollow fiber filter unit 28 to be periodically exchanged with the other one, the knotting-wire type filter unit 6 is used at a time of exchanging the hollow fiber filter unit 28, whereas the hollow fiber filter unit 28 is utilized at a time other than exchanging time. Namely, the cladding adheres to the hollow fiber filter even if the filter is not used for the actual driving water filtering operation, and accordingly, it is necessary to periodically exchange the hollow fiber filter even if not used, resulting in the troublesome workings and demerit in economy as well as increasing of the waste. On the other hand, the knotting-wire type filter can be used for a long time and only a short time is required for the exchanging of the hollow fiber filter. Based on the above technical facts, according to this embodiment, since the conventional knotting-wire type filter is utilized only at a time of the exchanging of the hollow filler filter, the generation of the waste can be reduced and the working of the operator can also be reduced as well as the economical merit. The arrangement of the two types of filters may prevent, reduce the simultaneous operation function problem of two filters in comparison with the usage of the same type. A fourth embodiment according to the present invention will be described hereunder with reference to FIG. 5, in which like reference numerals are added to elements or parts corresponding to the aforementioned embodiments and the description thereof is made simple or omitted. In the embodiment of FIG. 5, one of the suction filters 5 disposed on the inlet side of the driving water pump 4 shown in FIG. 8 is substituted with a hollow fiber filter unit 28. The driving water from the condensate demineralizer 3 passes the hollow fiber filter unit 28 just before the entrance into the driving water pump 4 to thereby remove foreign materials. According to this construction or arrangement, substantially the same effects and functions as those attained by the third embodiment can be achieved. In this embodiment, furthermore, since the driving water after the removal of the foreign materials enters the driving water pump 4, the biting of the foreign materials in the driving water pump 4 can be prevented, thus effectively maintaining the performance of the other elements or parts. A fifth embodiment according to the present invention will be described hereunder with reference to FIG. 6, in which like reference numerals are added to elements or parts corresponding to the aforementioned embodiments and the description thereof is made simple or omitted. In a case where only one suction filter 5 is disposed, the embodiment of FIG. 6 represents a case in which this one suction filter 5 disposed on the inlet side of the driving water pump 4 is substituted with a hollow fiber filter unit 28, and accordingly, in this embodiment, no suction filter is arranged. It is to be noted that according to this fifth embodiment substantially the same effects and functions as those attained by the former embodiment can be achieved. In the foregoing disclosures, the embodiments represented by FIGS. 3, 5 and 6 are mentioned with reference to FIG. 4, it is to be easily understood that the hollow fiber filter unit of FIG. 4 may be applicable to the embodiments of FIGS. 1 and 2 with minor but not substantial modification. It is also to be understood by persons in this art that the present invention is not limited to the described embodiments and other changes or modifications may be within the scope of the appended claims. |
claims | 1. A fuel assembly for a nuclear power boiling water reactor, comprising:a fuel channel extending in and defining a length direction of the fuel assembly and defining a central fuel channel axis extending in said length direction;fuel rods positioned such that they are surrounded by said fuel channel, each fuel rod having a central fuel rod axis extending substantially in said length direction; andwater channels positioned such that they are surrounded by said fuel channel, the water channels being configured and positioned for, during operation, allowing non-boiling water to flow through the water channels, each water channel having a central water channel axis extending substantially in said length direction,wherein said fuel rods comprise a first group of fuel rods and a second group of fuel rods,wherein each fuel rod in said first group is a full length fuel rod that extends from a lower part of the fuel assembly to an upper part of the fuel assembly,wherein each fuel rod in said second group is a part length fuel rod that extends from said lower part of the fuel assembly and upwards, but does not reach as high up as said full length fuel rods,wherein said water channels of the fuel assembly comprise at least three and no more than three water channels, each of which has a cross-sectional area that is at least twice as large as the average cross-sectional area of the fuel rods,wherein the three water channels are positioned with no further water channel having its central axis closer to the central fuel channel axis than the central water channel axis of each the three water channels, andwherein there are at least five second group fuel rods positioned with their central fuel rod axes being closer to the central fuel channel axis than any of the water channel axes of the water channels of the fuel assembly. 2. A fuel assembly according to claim 1, wherein no first group fuel rod is positioned with its central fuel rod axis closer to the central fuel channel axis than the central fuel rod axis of any of said at least five second group fuel rods. 3. A fuel assembly according to claim 1, comprising at least seven second group fuel rods positioned with their respective central fuel rod axes closer to the central fuel channel axis than any of the water channel axes of the water channels. 4. A fuel assembly according to claim 3, wherein no first group fuel rod is positioned with its central fuel rod axis closer to the central fuel channel axis than the central fuel rod axis of any of said at least seven second group fuel rods. 5. A fuel assembly according to claim 3, wherein each second group fuel rod has a length that is less than 0.50 times the length of said first group fuel rods. 6. A fuel assembly according to claim 5, wherein no fuel rod longer than 0.50 times the length of said first group fuel rods is positioned with its central fuel rod axis closer to the central fuel channel axis than the central fuel rod axis of any of said second group fuel rods. 7. A fuel assembly according to claim 1, wherein each second group fuel rod has a length that is less than 0.50 times the length of said first group fuel rods. 8. A fuel assembly according to claim 7, wherein no fuel rod longer than 0.50 times the length of said first group fuel rods is positioned with its central fuel rod axis closer to the central fuel channel axis than the central fuel rod axis of any of said second group fuel rods. 9. A fuel assembly according to claim 1, wherein the cross-sectional area of each water channel is between 3.0 and 10.0 times the cross-sectional area of each one of said fuel rods. 10. A fuel assembly according to claim 9, wherein the cross-sectional area of each water channel is between 4.0 and 8.0 times the cross-sectional area of each one of said fuel rods. 11. A fuel assembly according to claim 1 wherein the fuel assembly has fuel rods of different cross-sectional areas and wherein the cross-sectional area of each water channel is between 3.0 and 10.0 times the average cross-sectional area of the fuel rods. 12. A fuel assembly according to claim 11 wherein the fuel assembly has fuel rods of different cross-sectional areas and wherein the cross-sectional area of each water channel is between 4.0 and 8.0 times the average cross-sectional area of the fuel rods. 13. A fuel assembly according to claim 1, wherein each of said at least three and no more than three water channels has a circular cross-section, at least in the lower part of the fuel assembly where the second group fuel rods are arranged. 14. A fuel assembly according to claim 1, wherein the fuel assembly comprises no more than 19 fuel rods having a distance between the central fuel rod axis and the central fuel channel axis less than the distance between the central water channel axis of at least one of said three water channels and the central fuel channel axis. 15. A fuel assembly according to claim 14, wherein the fuel assembly comprises no more than 16 fuel rods having a distance between the central fuel rod axis and the central fuel channel axis is less than the distance between the central water channel axis of at least one of said three water channels and the central fuel channel axis. 16. A fuel assembly according to claim 15, wherein the fuel assembly comprises no more than 14 fuel rods having a distance between the central fuel rod axis and the central fuel channel axis is less than the distance between the central water channel axis of at least one of said three water channels and the central fuel channel axis. 17. A fuel assembly according to claim 1, wherein the fuel assembly comprises a substantially regular pattern of fuel rod positions, wherein each one of said three water channels is positioned such that it replaces four fuel rods in this substantially regular pattern. 18. A fuel assembly according to claim 1, wherein the fuel assembly comprises 65-160 fuel rods. 19. A fuel assembly according to claim 18, wherein the fuel assembly comprises 100-120 fuel rods. 20. A fuel assembly according to claim 19, wherein the fuel assembly comprises 105-113 fuel rods. 21. A fuel assembly according to claim 20, wherein the fuel assembly comprises 109 fuel rods. 22. A fuel assembly according to claim 1, wherein the fuel assembly further comprises 2-8 additional second group fuel rods, each of which has a length of between 0.59 and 0.79 times the length of said first group fuel rods. 23. A fuel assembly according to claim 22, wherein the fuel assembly further comprises 4-6 additional second group fuel rods, each of which has a length of between 0.59 and 0.79 times the length of said full length fuel rods. 24. A fuel assembly according to claim 1, wherein the fuel assembly comprises at least 70 full length fuel rods. 25. A fuel assembly according to claim 24, wherein the fuel assembly comprises at least 80 full length fuel rods. 26. A fuel assembly according to claim 25, wherein the fuel assembly comprises at least 90 full length fuel rods. 27. A fuel assembly according to claim 1, further comprising:a lower tie plate, positioned below the fuel rods, wherein a lower end of each of said three water channels is attached to said tie plate;an upper lifting device positioned above the fuel rods and including a handle for gripping and lifting a bundle of fuel rods;a plurality of spacer grids for holding the fuel rods, at least most of the spacer grids being attached to said three water channels; andattachment rods attached at their lower ends to the upper part of said three water channels and at their upper ends to said upper lifting device. |
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claims | 1. A method of producing a useful short lived radioisotope for medical or industrial applications from a first target isotope, the method comprising the steps of:providing a first buffer region around a neutron source for providing a first reduction in neutron energy by inelastic scattering;providing an activation region around the first buffer region, said activation region being made of heavy elements of lead and/or bismuth;distributing a material containing said first target isotope throughout the whole volume of the activation region, the inner buffer region and the neutron source being devoid of said first target isotope;activating the neutron source to emit a neutron flux such that neutrons of said neutron flux are captured by the first target isotope to produce said useful short-lived radio-isotope for medical or industrial applications; andrecovering said useful short-lived radioisotope from the exposed material for use in medical or industrial applications;wherein multiple elastic collisions between the neutrons in the neutron flux and the heavy elements in the activation region result in an enhanced neutron flux in the activation region; and a rate of progressive decrease in neutron energy such that neutron capture efficiency in said first target isotope is enhanced by resonance neutron capture. 2. A method according to claim 1, further comprising the step of providing a neutron moderator surrounding the activation region where the exposed material is distributed. 3. A method according to claim 2, further including the step of providing a second buffer region, made of said heavy elements free of the exposed material, located between the moderator and the activation region where the exposed material is distributed. 4. A method according to claim 2, wherein the moderator is made of carbon or deuterated water. 5. A method according to claim 4, wherein the moderator is made of carbon, and has a thickness of the order of 5 to 10 cm. 6. A method according to claim 1, wherein the neutron source consists of a central region of the lead and/or bismuth medium, which is bombarded with a high-energy charged particle beam to produce neutrons by spallation. 7. A method according to claim 6, wherein the lead and/or bismuth of said central region is in liquid phase, and is circulated by natural convection along a circuit including a heat exchanger and an auxiliary heater. 8. A method according to claim 1, wherein the neutron source consists of a beryllium or lithium target bombarded with a charged particle beam. 9. A method according to claim 1, wherein the neutron source is a radioactive source. 10. A method according to claim 1, wherein the neutron source consists of a spallation target bombarded with a high-energy charged particle beam. 11. A method according to claim 1, wherein the exposed material comprises 127I as said first isotope, which produces the useful radio-isotope 128I by capturing neutrons from the flux. 12. A method according to claim 11, wherein the exposed material is an iodine compound to be administered to patients after the neutron exposure. 13. A method according to claim 1, wherein the exposed material comprises 98Mo as said first isotope, which produces 99Mo by capturing neutrons from the flux, said 99Mo being allowed to decay into the useful radio-isotope 99mTc. 14. A method according to claim 13, wherein the exposed material comprises a phosphomolybdate complex salt which, after the neutron exposure, is absorbed in an alumina matrix from which the 99mTc is extracted after the decay of a substantial portion of the 99Mo. 15. A method according to claim 1, wherein the exposed material comprises 130Te as said first isotope, which produces 131Te by capturing neutrons from the flux, said 131Te decaying into the useful radio-isotope 131I. 16. A method according to claim 15, wherein the exposed material comprises metallic tellurium, which is melted after the neutron exposure so as to volatilise the iodine content thereof. 17. A method according to claim 1, wherein the exposed material comprises a fissile element as said first isotope, which produces fission fragments by capturing neutrons from the flux, said useful isotope being a radio-isotope extracted from said fission fragments. 18. A method according to claim 1, wherein the exposed material comprises 124Xe as said first isotope, which produces 125Xe by capturing neutrons from the flux, said 125Xe decaying into the useful radio-isotope 125I. 19. A method according to claim 1, wherein the exposed material comprises a semiconductor material, and the useful isotope is a doping impurity within said semiconductor material, which is obtained from neutron captures by a first isotope of the semiconductor material. 20. A method according to claim 19, wherein the semiconductor material consists of silicon, with 30Si as said first isotope producing 31Si by capturing neutrons from the flux, said 31Si decaying into 31P as an electron-donor doping impurity. 21. A method according to claim 19, wherein the semiconductor material consists of germanium, with 70Ge as said first isotope producing 71Ge by capturing neutrons from the flux, said 71Ge decaying into 71Ga as an electron-acceptor doping impurity, and also with 74Ge producing a smaller amount of 75Ge by capturing neutrons from the flux, said 75Ge decaying into 75As as an electron-donor doping impurity. |
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claims | 1. A gravitational wave generating device comprising: a plurality of target nuclei in a constrained state, a source of submicroscopic particles directed at the target nuclei, a computer-controlled logic system operatively connected to the particle source for selectively propelling the particles toward the target nuclei to cause products of a nuclear reaction to be emitted from the nuclei, and a containment system for aligning the products of the nuclear reaction such that the products move in approximately the same direction, produce a third time derivative of the motion of the target nuclei reacting to the emitted products of the nuclear reaction and thereby generate gravitational waves in that direction. 2. A device according to claim 1 in which the plurality of target nuclei are contained in a superconducting medium. claim 1 3. A device according to claim 1 in which the plurality of target nuclei comprises a fluid. claim 1 4. A device according to claim 3 wherein the fluid includes electrons. claim 3 5. A device according to claim 3 in which the fluid is a superconducting fluid. claim 3 6. A device according to claim 1 in which the plurality of target nuclei comprises a gas. claim 1 7. A device according to claim 1 in which the plurality of target nuclei are constrained in an electromagnetic field. claim 1 8. A device according to claim 7 in which the electromagnetic field is external to the plurality of target nuclei. claim 7 9. A device according to claim 7 in which the electromagnetic field is ferromagnetic. claim 7 10. A device according to claim 7 in which the electromagnetic field comprises intermolecular forces. claim 7 11. A device according to claim 1 in which the plurality of target nuclei are aligned in a spin-polarized state. claim 1 12. A device according to claim 1 in which the source of particles for producing nuclear-reaction products is a pulsed particle beam. claim 1 13. A device according to claim 12 in which the particles comprising the particle beam are photons. claim 12 14. A device for generating gravitational waves comprising a source of products of nuclear reactions under the control of a computer-controlled logic system to produce a third time derivative of the motion of energizable elements and thereby generate gravitational waves. 15. A gravitational wave generating device comprising: a plurality of target energizable elements, a plurality of energizing elements that act on the energizable elements, and a computer controlled logic system operatively connected to the energizing elements to control the action of the energizing elements so as to produce a third time derivative of the motion of the energizable elements or a jerk and thereby generate gravitational waves. 16. A device according to claim 15 in which the energizable elements are molecules. claim 15 17. A device according to claim 15 in which the energizable elements are atoms. claim 15 18. A device according to claim 15 in which the energizable elements are atomic nuclei. claim 15 19. A device according to claim 15 in which the energizable elements are nuclear particles. claim 15 20. A device according to claim 19 in which the nuclear particles are electrons. claim 19 21. A device according to claim 15 in which the energizing elements are an anisotropic particle beam. claim 15 22. A device according to claim 21 in which the beam particles collide with the energizable elements and produce a third time derivative of the motion of the energizable elements and generate gravitational waves. claim 21 23. A device according to claim 22 in which the beam particles collide with the energizable elements to produce a nuclear reaction that causes the ejection of nuclear reaction products that result in a third time derivative of the motion of the energizable elements. claim 22 24. A device according to claim 15 in which the energizing elements are an isotropic particle beam. claim 15 25. A device according to claim 15 in which the energizing elements create a multiquantum vibrational event for the energizable elements on a subpicosecond time scale and generate gravitational waves. claim 15 26. A device according to claim 15 in which the energizing elements are microwaves. claim 15 27. A device according to claim 15 in which the energizing elements are one or more magnetic fields. claim 15 28. A device according to claim 15 in which the energizing elements are one or more electric fields. claim 15 29. A device according to claim 15 in which the energizing elements move in sequence to define a gravitational-wave front and energize the energizable elements in sequential order to generate and accumulate gravitational-wave energy as the gravitational-wave front progresses. claim 15 30. A device according to claim 29 in which the gravitational waves comprising the wave front are coherent. claim 29 31. A device according to claim 15 in which the energizing elements are photons of a laser. claim 15 32. A device according to claim 15 in which the energizing elements are electrons. claim 15 33. A device according to claim 15 in which the energizing elements are protons. claim 15 34. A device according to claim 15 in which the energizing elements are neutrons. claim 15 35. A device according to claim 15 in which the energizing elements are nuclear particles. claim 15 36. A device according to claim 15 in which the energizing elements are atomic nuclei. claim 15 37. A device according to claim 15 in which the energizing elements are molecules. claim 15 38. A device according to claim 37 in which the molecules are ionized. claim 37 39. A device according to claim 15 , in which the energizing elements are current-carrying coils. claim 15 40. A device according to claim 15 , in which the energizable elements are one or more permanent magnets. claim 15 41. A device according to claim 40 in which the permanent magnets are submicroscopic. claim 40 42. A device according to claim 15 , in which the energizable elements are one or more electromagnets. claim 15 43. A device according to claim 42 in which the electromagnets are submicroscopic. claim 42 44. A device according to claim 15 , in which the energizing elements are current-carrying electrical conductors. claim 15 45. A device according to claim 15 , in which the energizable elements are current-carrying electrical conductors. claim 15 46. A device according to claim 15 in which the energizable elements are energized in a pattern in order to achieve directivity in gravitational wave transmission. claim 15 47. A device according to claim 46 in which the directivity is changed over time in order to control the direction of the gravitational wave transmissions. claim 46 48. A device according to claim 46 in which the energizing elements are energized in a pattern that will transmit gravitational waves to a radiating gravitational wave transmitter in order to establish a GW communications source. claim 46 49. A device according to claim 46 in which the pattern produces constructive interference among some of the gravitational waves. claim 46 50. A device according to claim 46 in which the pattern produces destructive interference among some of the gravitational waves. claim 46 51. A device according to claim 15 in which the energizable elements are harmonic oscillators. claim 15 52. A device according to claim 15 in which the energizable elements are capacitors. claim 15 53. A device according to claim 15 in which the energizable elements are disposed in a spherical array. claim 15 54. A device according to claim 53 in which the spherical array comprises piezoelectric crystals spread evenly over the surface of a sphere. claim 53 55. A device according to claim 53 in which the energizable element comprise a spherical piezoelectric crystal or crystals. claim 53 56. A device according to claim 55 in which actuating electrodes are spread evenly over the surface of the piezoelectric crystals and operatively connected to a power source controlled by a computer-controlled logic system. claim 55 57. A device according to claim 15 in which a refractive medium concentrates or focuses the gravitational waves emitted by the gravitational wave generator. claim 15 58. A device according to claim 15 , in which the energizable elements are piezoelectric crystals. claim 15 59. A device according to claim 15 , in which the energizable elements are nanomachines. claim 15 60. A device according to claim 59 in which the nanomachines are harmonic oscillators. claim 59 61. A device according to claim 59 in which the nanomachines are nanomotors. claim 59 62. A device according to claim 59 in which the nanomachines are solenoids. claim 59 63. A device according to claim 59 in which the nanomachines are microelectromechanical systems (MEMS). claim 59 64. A device according to claim 15 in which the energizing elements are antiprotons. claim 15 65. A device according to claim 15 in which the energizable elements are antiprotons. claim 15 66. A device according to claim 15 in which the energizable elements are enveloped in a dielectric. claim 15 67. A device according to claim 66 in which the dielectric has a spherical form. claim 66 68. A device according to claim 15 in which the energizing elements are sources of electromagnetic radiation. claim 15 69. A device according to claim 15 in which the energizable elements are submicroscopic particles. claim 15 70. A device according to claim 15 in which the computer-controlled logic system is a modulator. claim 15 71. A device according to claim 15 in which the energizable elements are maintained in a state of superconductivity. claim 15 72. A device according to claim 15 in which the computer-controlled logic system is a vehicle trajectory or orbit determination processor. claim 15 73. A gravitational wave detection device comprising gravitational-wave collector elements that are interrogated by a computer-controlled logic system according to an expected arrival time of the crests of a gravitational wave of a predetermined gravitational wave frequency and phase in order to be a tuned gravitational wave receiver. 74. A device according to claim 73 in which the interrogations continue as the gravitational wave phase is determined and locked on by a control computer. claim 73 75. A device according to claim 73 in which the collector elements are transducers. claim 73 76. A device according to claim 75 in which the transducers are parametric transducers. claim 75 77. A device according to claim 75 in which the transducers measure the curvature of the spacetime continuum. claim 75 78. A device according to claim 73 in which the collector elements are capacitors. claim 73 79. A device according to claim 73 in which the collector elements are harmonic oscillators. claim 73 80. A device according to claim 73 in which signals from the collector elements can be measured by a superconducting quantum interference device (SQUID). claim 73 81. A device according to claim 73 in which the signal from the collector elements are sensed using quantum non-demolition (QND) techniques. claim 73 82. A device according to claim 73 in which the collector elements are interrogated in a pattern according to an expected incoming gravitational wave direction in order to achieve directivity in GW reception. claim 73 83. A device according to claim 73 in which the directivity is changed over time in order to scan for gravitational wave transmissions. claim 73 84. A device according to claim 73 in which the collector elements are disposed in a spherical array. claim 73 85. A device according to claim 84 in which the spherical array of collector element comprises a plurality of piezoelectric crystals spread evenly over the surface of a sphere. claim 84 86. A device according to claim 84 in which the collector element comprise spherical piezoelectric crystals. claim 84 87. A device according to claim 86 in which actuating electrodes are spread evenly over the surface of the piezoelectric crystals and operatively connected to a computer-controlled logic system. claim 86 88. A device according to claim 73 in which the collector elements are submicroscopic. claim 73 89. A device according to claim 73 in which the tuned gravitational wave receiver receives gravitational waves refracted by a medium positioned in front of the gravitational-wave receiver. claim 73 90. A device according to claim 89 in which the medium is a superconducting medium. claim 89 91. A device according to claim 89 including a lens for concentrating or focusing the gravitational waves. claim 89 92. A device according to claim 89 including a series of gravitational-wave refracting media for concentrating or focusing the gravitational waves. claim 89 93. A device according to claim 73 in which the collector elements are maintained in a state of superconductivity. claim 73 94. A gravitational wave communications device comprising: a gravitational wave generator for producing gravitational waves having a particular frequency and amplitude as determined by the frequency and amplitude of the jerks of the masses comprising the gravitational wave generator, a modulator connected to the generator for imparting information to the gravitational waves by modifying their frequency and amplitude, a computer-controlled logic system for controlling the frequency and amplitude of the jerks, a detector for receiving the modulated gravitational waves having a particular frequency and amplitude, and a demodulator controlled by a computer-control logic system for extracting the information from the frequency and amplitude of gravitational waves and delivering it to a presentation device. 95. A gravitational wave communications device comprising: a plurality of target nuclei aligned in a constrained state, a source of submicroscopic particles directed at the target nuclei, a computer-controlled logic system operatively connected to the particle source for selectively propelling the particles toward the target nuclei to produce a nuclear reaction, a containment system for aligning the products of the nuclear reaction such that the particles move in approximately the same direction, produce a third time derivative in the motion of the target nuclei and thereby generate gravitational waves, and a transmitter operatively connected to the containment system for selecting the number of particles propelled at any given time to modulate the gravitational waves. 96. A device according to claim 95 wherein the transmitter includes a modulator. claim 95 97. A device according to claim 96 in which the modulator imparts information to the gravitational waves by selecting their frequency and amplitude. claim 96 98. A device according to claim 97 including a detector at a remote location for receiving the modulated gravitational waves. claim 97 99. A device according to claim 98 including a demodulator connected to the detector. claim 98 100. A device according to claim 99 including a presentation device connected to the demodulator. claim 99 |
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053012119 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIGS. 1 and 2 are schematic representations of the use of the so-called Linear Magnatron Sputtering technique, available from Surface Solutions, Inc., as adapted for implementing the present invention. In the preferred embodiment, a nuclear fuel assembly component such as fuel rod cladding or a control rod guide tube 10, made from zircaloy, is coated substantially uniformly along the entire inside surface, over substantially the full length of the tube. For illustrative purposes, the aspect ratio of the component tube 10 as shown in the Figures, has been substantially shortened relative to the typical tube dimensions of less than 0.5 inch ID and a length in excess of 12 ft. The component tube 10 is fixtured such that first and second end plugs 12, 14 seal the tube interior. One end plug 14 has a vacuum port 16 through which a vacuum pump 18 can evacuate the content of the tube. The working gas port 28 may be situated in either end plug. Each end plug includes means, such as first and second mounts 20,22, for supporting a source tube 100 of wear or corrosion resistant material having a smaller outer diameter than the inner diameter of the component tube, coaxially within the component tube, thereby defining a cylindrical annulus 24 between the tubes. After evacuation, an inert working gas such as argon from source 26 is backfilled in the annulus 24 to a pressure sufficient to sustain a plasma discharge. The source tube 100 in the normal implementation of the present invention, would be a homogeneous tube of the material which is desired to be coated on the inside surface, or substrate 30, of the cladding tube 10. In the preferred embodiment, a chemical reaction occurs in the space 24 between the source material from tube 100 and a reactive gas, in a manner to be described more fully below, whereby an oxide, nitride, or carbide is coated on the substrate. The source tube 100 can be at least the same length as the cladding tube 10. A power supply 32 with negative lead 34 is connected through the connector mount or otherwise, to the source tube 100, and the positive lead 36 is connected to the substrate 30, such that the source tube serves as a cathode and the substrate serves as an anode. A plasma 37 consisting of positive argon ions and electrons is established in the annular space 24, with the positive ions 38 bombarding the cathode 100 with sufficient energy to vaporize surface atoms from the source tube 100 onto the substrate 30. Because the source material is passed into the vapor phase 40 by a mechanical process (momentum exchange) rather than a chemical or thermal process, virtually any material is a candidate for coating. Thus metals, glasses and other ceramics having desirable wear and corrosion-resistant properties can be utilized. The plasma in the annular space can be established by any known means, such as direct current discharge when sputtering metals, but in order to improve the efficiency, magnatron techniques are applied to confine and shape the plasma. In the Linear Magnatron Sputtering technique available from Surface Solutions, Inc. a circumferential magnetic field 42 is established around the source tube 100 by running high currents from current source 44 axially through a copper tube 46 centered within the source tube and cooled by coolant 50, as shown in FIG. 2. This achieves a uniform rate of material evaporation along the length of the source tube 100, because of a constant plasma thickness that is uniformly excited over the whole length of the source tube. This results from the plasma drift current in the plasma surrounding the source tube, running in a direction parallel to the axis of the source tube 100. This is far superior to cylindrical post sputtering schemes, which require that the drift current run in a closed loop. In the Linear Magnatron Sputtering System of the present invention, the drift current may be boosted with an enhancer device shown generally at 48, to a very high level at one end of the source tube. It should thus be appreciate that to the extent a plasma of constant thickness can be maintained in uniform excitation over the full length of the source tube 100, the radially projecting source atoms 40 will coat the full surface 30 of the cladding tube substantially uniformly. Under many circumstances, satisfactory results are obtained if at least about ten feet along a twelve foot tube, is coated. It should also be appreciated that in the embodiment described in FIG. 1, the component tube 10 itself serves as the boundary of the evacuation chamber 24. An alternative embodiment would encapsulate the tube 10 within an outer vacuum chamber (not shown), in which case the fixturing and end plug arrangements would have analogous counterparts in the walls of the vacuum chamber. Once initiated, the flux of sputtered material leaving the source tube 100 will continue until a substantially uniform coating of, for example, a thickness of 0.0002 inch is achieved, whereupon the process is stopped, the end plugs removed, and the next tube fixtured for restarting the coating process. FIG. 3 shows an alternative embodiment of the process, and the coated tube resulting from such processes. In FIG. 3, the cladding or guide tube 200 has a non uniform but prescribed distribution of coating material, as a result of a commensurate non uniform, but prescribed distribution of material on source tube 300. For example, if component 200 is a control rod guide tube, the upper or left portion as shown in FIG. 3a can be coated at 210 over only a limited axial distance, e.g., less than one half the guide tube length, extending to just below the "withdrawn" portion of the control rod tips in the nuclear reactor. The source tube 300 has a correspondingly limited region of source material 310. Thus, the coating can be limited to partial lengths within the rod by shortening the length of source tube. As mentioned above, the process according to the present invention is amenable to the sputter coating of a wide variety of potentially useful materials. Some of these materials cannot readily be sputtered from a homogeneous source tube, but rather can be formed either chemically in the inert gas via chemical reaction, or at the internal surface of the component tube. The coatings are thus applied by analogy to reactive chemical deposition processes. A source 52 of reaction gas including nitrogen, oxygen, or carbon and associated plug port 54 are shown in FIG. 1 for this purpose. Table 1 lists a variety of metals and metallic compounds that can be sputtered in accordance with the present invention: TABLE 1 ______________________________________ Metals and Metallic Compounds ______________________________________ ZrN TiN CrN HfN TiAlVN TaN ______________________________________ Table 2 is a representative list of ceramic materials including glasses that are usable with the present invention: TABLE 2 ______________________________________ Ceramics and Glasses ______________________________________ Zr.sub.2 O.sub.3 Al.sub.2 O.sub.3 TiCN TiC CrC ZrC WC Calcium Magnesium aluminosilicate Sodium Borosilicate Calcium Zinc borate ______________________________________ |
abstract | In a nuclear fuel storage rack including a plurality of rack cells (1) configured to house a nuclear fuel assembly, the rack cell (1) includes a plurality of plate members (10 to 13) containing a radiation absorption material and stood to form a nuclear fuel housing space (8) configured to house the nuclear fuel assembly, and a fastening unit (14) configured to fasten the plurality of plate members (10 to 13). Each of the plate members (10 to 13) includes projections (16) protruding outward in a lateral direction from one side end and the other side end extending in an upward/downward direction (T1), and concave sections (17) formed at one side end side and the other side by the projections (16). The projections (16) protruding outward from outer surfaces of the plate members (10 to 13) in the lateral direction are fastened by the fastening unit (14). |
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048204722 | claims | 1. A fuel rack for storing nuclear fuel assemblies in a nuclear fuel-storage pool having a floor; the said fuel rack comprising: a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, said base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams vertically spaced from the first network forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each said cell being positioned in a corresponding pair of said aligned polygonal openings, each said cell being open at both ends with a guiding funnel at its upper end, and said cells being positioned over said flow openings in said base-plate means, thereby to permit flow of coolant through said cells; spaced outwardly direction projections on said outer surfaces of the sides of said cells near the tops and bottoms of the sides thereof, each said cell together with said projections thereon being sized to be received within a corresponding said pair of aligned polygonal openings in which said each cell is positioned; means fixedly securing said projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design; neutron absorbing material mounted on the outer surfaces of the sides of at least some of said cells between said spaced top and bottom projections, said neutron absorbing material having a thickness no greater than the outward extent of said projections; and leveling means for said base structure located between the fuel storage pool floor and said base structure for adjusting the base-plate means and cells thereon to a level condition. each of said pads including a base member having vertically adjustable means therein arranged to engage the underside of said plate and thereby provide, upon adjustment, a horizontal surface to help assure vertical alignment of the cell walls. so that when the foot is adjusted vertically, the frame and frame arms move relative to the pool floor to achieve leveling of the plate to assure vertical alignment of the cell walls. a spherical surface on said pedestal; and a complementary spherical surface on the bottom of said foot which engages the pedestal thus permitting the foot to move on said spherical surface relative to the pedestal and assume a vertical position when the pedestal rests on a non-horizontal surface. indentations on the upper end of said foot engageable by a tool which permits the foot to be adjusted to leveling positions. said leveling pad including a pedestal mounted on the pool floor; a vertically adjustable foot having one end in said pedestal and the other end positioned for access through an opening in the base plate; a plate support member having radially extending arms arranged to contact the underside of said base plate, said member having a central threaded opening into which said foot is threaded so that upon rotation of the foot in the pedestal, the support member is caused to move vertically and thus adjust the leveling position of the base plate to assure vertical alignment of the cell walls. means associated with each cell for providing a continuous structural support for said material. a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, said baseplate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each said cell being positioned in a corresponding pair of said aligned polygonal openings, each said cell being open at both ends with a guiding funnel at the upper end, and said cells being positioned over said flow openings in said base-plate means, thereby to permit flow of coolant through said cells; spaced, outwardly directed, projections on said outer surfaces of the sides of said cells near the tops and bottoms of the sides thereof, each said cell together with said projections thereon being sized to be received within a corresponding of said pair of aligned polygonal openings in which said cells are respectively positioned; and means fixedly securing said projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design; neutron absorbing material mounted on the outer surfaces of the sides of at least some of said cells between said spaced top and bottom projections, said neutron absorbing material having a thickness no greater than the outward extent of said projections; and leveling means for said base structure, located between the fuel storage pool floor and said base structure, for adjusting the base-plate and the cells thereon to a level condition; said leveling means for the base structure including: multiple leveling pads beneath said base structure, each of said pads including a base member having vertically adjustable means therein arranged to engage the underside of said base structure and thereby provide, upon adjustment, a horizontal surface to help ensure vertical alignment of the cells, said vertically adjustable means including a frame having upstanding arms which engage the underside of the base structure and an adjustable foot arranged to coact with said frame to adjust vertically the base structure to assure vertical alignment of the cell walls, said frame having a central opening and said adjustable foot being screw threaded into said opening so that when said foot is adjusted vertically, said frame and frame arms move relative to the pool floor to achieve leveling of the base structure and also to assure vertical alignment of the cell walls; the said nuclear fuel rack also including: means, connected to the foot, actuable to permit the foot centerline to lie perpendicular to said base structure even though the floor of the pool on which said vertically adjustable means rests is uneven, said actuable means including: a pedestal to be positioned on the pool floor, a spherical surface on said pedestal, and a complementary spherical surface on the bottom of said foot which engages the pedestal on the spherical surface thereof thus permitting the foot to move on said spherical surface of said pedestal relative to the pedestal and assume a vertical position when the pedestal rests on a non-horizontal surface; said foot and said pedestal having coaxial openings to accommodate the upwardly projecting stud; and locking means on said foot and pedestal, cooperative with said stud, for locking said foot and pedestal in position on said floor. a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, said base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams vertically spaced from the first network forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each said cell being positioned in a corresponding pair of said aligned polygonal openings, each said cell being open at both ends with a guiding funnel at its upper end, and said cells being positioned over said flow openings in said base-plate means, thereby to permit flow of coolant through said cells; spaced outwardly directed projections on said outer surfaces of the sides of said cells near the tops and bottoms of the sides thereof, each said cell together with said projections thereon being sized to be received with a corresponding said pair of aligned polygonal openings in which said each cell is respectively positioned; means fixedly securing said projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design; neutron absorbing material mounted on the outer surfaces of the sides of said cells, said neutron absorbing material comprising sheet material covering substantially the full surface area of each side of each said cell, each said sheet on each side of each cell being enclosed on a wrapper plate, said plate being larger than said sheets and having its end edges bonded to the cell surfaces, said nuclear fuel rack also including means, associated with each cell, for providing a continuous structural support for said sheet material between said spaced top and bottom projections; and leveling means for said base structure located between the fuel storage pool floor and said base structure for adjusting the base-plate means and cells thereon to a level condition. a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, said base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings, a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each said cell being positioned in a corresponding pair of corresponding said aligned polygonal openings, each said cell being open at both ends, and said cells being positioned over said flow openings in said baseplate means, thereby to permit flow of coolant through said cells; spaced, outwardly directed, projections on said outer surfaces of the sides of said cells near the tops and bottoms of the sides thereof, said cells together with said projections thereon being sized to be received within said pair of corresponding aligned polygonal openings in which said cells are respectively positioned; and means fixedly securing said projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design; neutron absorbing material mounted on the outer surfaces of the sides of at least some of said cells between said spaced top and bottom projections, said neutron absorbing material having a thickness no greater than the outward extent of said projections; and leveling means for said base structure, located between the fuel storage pool floor and said base structure, for adjusting the base-plate and the cells thereon to a level condition; said leveling means for the base structure including: multiple leveling pads beneath said base structure, each of said pads including a base member having vertically adjustable means therein arranged to engage the underside of said base structure and thereby provide, upon adjustment, a horizontal surface to help ensure vertical alignment of the cells, said vertically adjustable means including a frame having upstanding arms which engage the underside of the base structure and an adjustable foot arranged to coact with said frame to adjust vertically the base structure to assure vertical alignment of the cell walls, said frame having a central opening and said adjustable foot being screw threaded into said opening so that when said foot is adjusted vertically, said frame and frame arms move relative to the pool floor to achieve leveling of the base structure also to assure vertical alignment of the cell walls. a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, said baseplate means supporting a first grid network of interlaced beams which form a multiplicity of polygonal openings; a second grid network of interlaced beams vertically spaced from said first network and forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, said sides of each said cell in transverse cross section forming a polygonal shape geometrically similar to the shape of said polygonal openings formed by said interlaced grid networks of beams, each said cell being positioned in a corresponding pair of said aligned polygonal openings over said flow openings in said base-plate means thereby to permit flow of coolant through said cell; spaced outwardly directed projections on said outer surfaces of all sides of each said cell near the tops and bottoms of the sides thereof, each said cell together with said projections thereon being sized to be received within a corresponding of said pair of aligned polygonal openings in which said each cell is positioned; and means fixedly securing said projections on each cell on all sides to the adjacent beams defining the openings through which said each cell passes. a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, said base-plate means supporting a first grid network of interlaced beams which form a multiplicity of polygonal openings; a second grid network of interlaced beams vertically spaced from said first network and forming a polygonal opening positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, said sides of each said cell in transverse cross section forming a polygonal shape geometrically similar to the shape of said polygonal openings formed by said interlaced grid networks of beams, each said cell being positioned in a corresponding pair of said aligned polygonal openings over said flow openings in said base-plate means thereby to permit flow of coolant through said cells; and means fixedly securing each said cell on all sides to the adjacent beams both in the first and second grid networks, which adjacent beams define the openings through which said each cell passes. 2. The fuel rack according to claim 1 wherein the ends of said first and second beams are welded to peripheral side plates which extend around the fuel rack module. 3. The fuel rack according to claim 1 wherein said outwardly directed projections comprise plates welded to the sides of each cell near the top and bottom ends thereof, said plates being of a thickness greater than the neutron absorbing material to permit the loading of the cell with the material attached thereto into the aligned square openings. 4. The fuel rack according to claim 1 wherein said outwardly directed projections include dimples formed in the cell walls, and the means securing the dimples to said beams comprise a weld made in the intercell space so that the dimpled cell material serves to absorb impact loads of the stored fuel assembly on the cell wall, thereby reducing potential damage to both the fuel rack and stored fuel assembly during seismic events. 5. the fuel rack according to claim 4 wherein the dimples project into the space between adjacent cells a distance greater than the outward extent of the neutron absorbing material to protect the material when the cell is loaded into the aligned square openings. 6. The nuclear fuel rack according to claim 1 wherein said base plate leveling means includes multiple leveling pads beneath said plate; 7. The nuclear fuel rack according to claim 6 wherein said vertical adjustable means includes a frame having upstanding arms which engage the underside of said base plate, and an adjustable foot arranged to coact with said frame to vertically adjust the base plate to assure vertical alignment of the cell walls. 8. The nuclear fuel rack according to claim 7 wherein said frame has a central opening and the adjustable foot is screw threaded into said opening; 9. The nuclear fuel rack according to claim 8 wherein means associated with said foot acts to permit the foot center line to lie perpendicular to the base plate even though the floor of the pool on which the adjustable means rests is uneven. 10. The nuclear fuel rack according to claim 9 wherein said means associated with said foot includes a pedestal adapted to be positioned on the pool floor; 11. The nuclear fuel rack according to claim 10 wherein the pedestal is permitted unrestricted movement on the pool floor. 12. The nuclear fuel rack according to claim 8 wherein the upper end of the adjustable foot is accessible through an opening in the base plate; and 13. The nuclear fuel rack according to claim 6 wherein leveling pads are positioned beneath the central area of said plate which supports the cells; 14. The nuclear fuel rack according to claim 1 wherein the neutron absorber material comprises sheet material which covers substantially the full surface area of each side of each cell; and 15. A fuel rack for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor on which an upwardly projecting stud is mounted; the said fuel rack comprising: 16. A fuel rack for storing nuclear fuel assemblies in a nuclear fuel-storage pool having a floor; the said fuel rack comprising: 17. The nuclear fuel rack according to claim 16 wherein each neutron absorber sheet material terminates short of the side edges of each cell and is substantially the same length as the active portion of a fuel assembly adapted for positioning in the cell. 18. A fuel rack for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor, the said fuel rack comprising: 19. A free-standing fuel rack for storing nuclear fuel assemblies in a nuclear fuel-storage pool; said fuel rack comprising: 20. The free-standing fuel rack according to claim 19 wherein neutron absorbing material is mounted on the outer surfaces of the sides of at least some of the cells between the spaced top and bottom projections, said neutron absorbing material having a thickness no greater than the outward extent of said projections. 21. The free-standing nuclear fuel rack according to claim 19 characterized by that the neutron absorber material comprises sheet material which covers substantially the full surface area of each side of each cell and further characterized by means, associated with each cell, for providing a structural support for said material. 22. A free-standing fuel rack for storing nuclear fuel assemblies in a nuclear fuel-storage pool; said fuel rack comprising: 23. The free-standing fuel rack of claim 22 wherein the polygonal openings along the periphery of said fuel rack, both in the first and second grid networks, have open gaps characterized by that said fuel rack includes upper and lower side plates extending along the periphery of said rack, across said open gaps and further characterized by means for fixedly securing each cell to the parts of the side plates extending across the gaps of the peripheral polygonal openings through which said each cell passes. |
053496183 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates generally to boiling water nuclear reactors (BWRs) and more particularly to nuclear fuel assemblies and the fuel rods used in BWRs for electrical power generation. 2. Description of Related Art The generation of heat energy through fission of nuclear fuel in a nuclear reactor is well known. The nuclear fuel is located in the core of the reactor. Typically, the core of a light-water reactor (LWR) contains a plurality of fuel assemblies that each contain a plurality of fuel rods. The fuel rods are arranged with their axes along the vertical direction. Typically, a fuel rod consists of a metal cylindrical cladding that contains a stack of fuel pellets. The fuel pellet material, the size of the fuel pellets, the cladding material and the orientation of the fuel pellets within the fuel rod are all well known for commercial LWRs used in electric power production. The radial spacing between fuel rods is filled with water, which flows in the axial direction from the bottom of the core to its top. The water carries away heat from the reactor core that is generated by fission reactions within the fuel rods. Hence, the water acts as the core coolant. The water also slows down, i.e., moderates, neutrons which are emitted at high velocities from the fission reactions. Hence, the water acts also as a neutron moderator. The moderation of high velocity fission neutrons occurs as a result, primarily, of collisions between the neutrons and the nuclei of hydrogen atoms in the water. Consequently, the degree of moderation is a function of the number of hydrogen atoms per unit volume, for example, per cubic centimeter. As the water is heated, the water expands and so the number of hydrogen atoms per cubic centimeter decreases. When the water boils and is converted to steam, the number of hydrogen atoms per cubic centimeter decreases even further. Consequently, a region of a reactor core in which the steam occupies a significant fraction of the volume available for the coolant is undermoderated. That is, neutrons in such a region are not sufficiently moderated. The moderation of a fission neutron, i.e, the lowering of the energy of the fission neutron, increases the probability that the neutron is absorbed in the nuclear fuel so as to cause another fission reaction which in turn releases additional high energy neutrons. When neutrons from one fission reaction cause, on the average, one other fission reaction to occur, a chain reaction is established. To generate power, a chain reaction must be established and maintained in a LWR. The probability for achieving a chain reaction is sometimes measured by a multiplication constant and is denoted by K. In LWRs, the neutron moderation provided by the coolant makes it possible to obtain a nuclear chain-reaction, i.e., a K of one, using nuclear fuel having a relatively small concentration of a fissile isotope. Typical LWR fuel contains uranium in which the concentration of the fissile isotope uranium-235 is enriched from the 0.71% found in natural uranium to approximately 3%. The remaining 97% of the uranium is the non-fissile isotope uranium-238. One objective in nuclear power reactor design is to lower the fuel enrichment because this reduces the nuclear fuel cost and enhances uranium resource utilization. The enrichment of the fuel loaded into the core is chosen so that the multiplication constant K is sufficiently higher than one so that the LWR can be operated for a significant period of time, usually about 12 to 18 months, without refueling. During this period of operation, sometimes referred to as a cycle, multiplication constant K is reduced to or near one as a result of depletion, i.e., burnup, of the fissile fuel and the accumulation of neutron absorbing fission products. The deviation of multiplication constant K from unity at any instant during the cycle, commonly referred to as reactivity and equal to (K-1)/K, is offset by insertion into the core of strong neutron absorbing materials. The strong neutron absorbing materials are introduced into the core in a variety of ways, e.g., via the insertion of control rods, via inclusion of burnable poisons in the fuel, via addition of boron to the water, or via a combination of these measures. The rate at which heat can be generated per unit length of a fuel rod, sometimes referred to as the linear-heat-rate, is limited by the ability of the water to carry this heat out of the core without having the fuel temperature and the cladding temperature exceed predetermined permissible values. The maximum permissible linear-heat-rate translates, in a given core, to the maximum power density, or power generated per unit volume of the core at the specific location under consideration. The power density in a LWR varies across the core both axially and radially. The power density tends to drop at the core periphery because there is an increased probability that fission neutrons will leak out of the core without causing another fission reaction. The power density also tends to drop in the vicinity of control rods. The control rods, which are used to regulate the chain-reaction, contain materials, such as boron, that have a large capture probability for moderated neutrons. The power density is usually depressed also in core regions which are undermoderated relative to other core regions. The depressions in power density are undesirable except, possibly, near the core periphery. The flatter the power density, i.e, the more the power density is constant across the core, the smaller the size of the core required to generate a given amount of electricity and consequently the more economical the reactor. Alternatively, the flatter the power density of a core of a given volume and power, the longer the residence time of the fuel in the reactor which also improves the reactor economics. The advantages of a flat power distribution have been widely recognized, but such distributions are limited by other factors, which are described more completely below. FIG. 1 illustrates a cross-sectional view of a portion of a typical boiling water reactor BWR core 20 containing a plurality of a rectilinear fuel assemblies 12 and a plurality of control rods 24. (A boiling water reactor is one type of a LWR.) Each fuel assembly 12 includes a square channel 21 that encloses an array of fuel rods and water rods. A gap 22 exists between all fuel assemblies 12. Gap 22 is filled with water and is called a "water gap." Gap 22 is used for the insertion of control rods 24, when necessary. The non-specified dimensions, materials, and other parameters, components and instruments associated with or part of the BWR of these and subsequent figures are conventional and are well-known to those skilled in the art. Examples of typical BWR fuel assemblies are depicted and described in U.S. Pat. No. 3,350,275, entitled "Reactor Fuel Assembly Device" issued to Venier et al. on Oct. 31, 1967; U.S. Pat. No. 3,466,226, entitled "Nuclear Fuel Element" issued to Lass on Sep. 9, 1969; U.S. Pat. No. 3,802,995, entitled "Nuclear Fuel assembly" issued to Fritz et al. on Apr. 9, 1974; and U.S. Pat. No. 4,664,882, entitled "Segmented Fuel and Moderator Rod" issued to Doshi on May 12, 1987, all of which are incorporated herein by reference in their entirety. Typical BWR fuel assemblies include a 7.times.7, an 8.times.8, a 9.times.9, or a 10.times.10 array of rods. For example, in an 8.times.8 array, the fuel assembly contains 62 fuel rods and two water rods. See for example, U.S. Pat. No. 3,802,995, entitled "Nuclear Fuel Assembly" issued to Fritz et al. on Apr. 9, 1974. A primary purpose of the water rods, which are typically in the interior of the fuel assembly, is to provide additional moderation. BWR fuel assembly 12 may also include all fuel rods. A drawback of a BWR fuel assembly with all fuel rods is that the central region of the fuel assembly is undermoderated relative to the fuel assembly periphery, as the water of water gaps 24 contributes to neutron moderation near this periphery. One consequence of this undermoderation is that the power density at the central part of the assembly is lower than at the periphery of the assembly. The resulting nonuniform radial power distribution across the fuel assemblies has a negative effect on the BWR economics. This is so because such a power distribution limits the total power output and the uranium fuel utilization of fuel assembly 12. While the water rods mitigate the undermoderation in the central region of fuel assembly 12, the water rods reduce the amount of fuel per assembly. Consequently, the total power that can be obtained from the BWR core and the fuel residence time in the BWR is limited by either the water rods or the nonflattened radial power distribution in a fuel assembly when the water rods are eliminated. Weitzberg proposed in U.S. Pat. No. 4,591,479, entitled "Boiling Water Reactor Fuel Bundle," and issued on May 27, 1986 to improve the moderation in the central region of a BWR fuel assembly by replacing fuel pellets in a certain number of fuel rods in the center region of the assembly by pellets made of zirconium hydride. However, this approach is less effective and more expensive than the use of water rods. In a BWR, the coolant, i.e., water, flows axially from the bottom of the core to the top of the core. As the water moves through the core, the water is heated by the heat generated in the fuel rods. In fact, the water starts to boil and to create steam before leaving the core. From the boiling initiation point upwards, the amount of liquid decreases and the amount of vapor, i.e., steam, increases. The larger the volume fraction occupied by the steam, the smaller becomes the number of hydrogen atoms per unit volume and the smaller becomes the moderation capability of the water. Thus, BWR cores are highly undermoderated in the upper region where steam is formed and the undermoderation increases as the volume of steam increases higher in the core region. As a result of the undermoderation in the upper regions of the core, the fission rate and, hence, the axial power density is highly asymmetric unless measures are taken to flatten the axial power distribution. A typical axial power distribution is illustrated in FIG. 2. The axial power density tends to peak in the lower part of the core and to strongly decline near the top of the core. If uncorrected, this non-uniform axial power shape can limit the overall reactor power output, as the peak power density can not exceed a given limit. The non-uniform axial power shape results in an uneven axial burn-up of the fuel. At the beginning-of-life of the fuel assembly, the rate of fuel consumption at the lower part of the assembly is significantly higher than at the upper part of the assembly. This non-uniform fuel consumption impairs overall fuel utilization. Moreover, near the end of a cycle, the depleted fuel at the bottom core results in a lower fission rate which in turn reduces the amount of heat transferred into the water at the lower part of the core. Thus, with time the onset of boiling moves higher up the core. As the steam volume fraction decreases, the reactivity in the upper part of the core becomes significantly higher than the reactivity at the lower part of the core. This reactivity imbalance can impair reactor safety as it reduces the effectiveness of the control and safety rods, which in a BWR enter the core from its bottom. Specifically, the highly non-uniform fuel consumption can reduce the cold shutdown reactivity margin as well as prolong the time it takes to quickly shut-down, i.e., scram, the reactor. Another drawback of the undermoderation in the upper region of a BWR core is that a significant fraction of the fission neutrons born in the region leak out of the core before being moderated. These relatively energetic neutrons damage components located in the vicinity of the core. A number of measures are presently being used in boiling water reactors to modify the axial power shape so as to minimize the disadvantages of the asymmetric axial power distribution. These measures include the use of control rods and of consumable neutron absorbing materials, also known as burnable poisons. Burnable poisons, such as gadolinium, are generally incorporated within the fuel rod unevenly, with the highest concentration in the lower part of the rod. Having a large probability of absorbing moderated neutrons, the burnable poison reduces the fission probability and, hence, suppresses the fission density. In addition, control rods are partially inserted into the core from its bottom, thereby absorbing moderated neutrons and thus suppressing fissions in this part of the core. Both of these methods of axial power shaping have an adverse effect on the fuel utilization in a BWR. If the number of neutrons absorbed in the burnable poisons and in the control rods were reduced, either the enrichment of the uranium loaded into the BWR could be reduced or the fuel residence time and burnup in the reactor could be increased. Both of these changes could improve the BWR economics. Moreover, both of these methods of axial power shaping do not eliminate the safety and neutron damage drawbacks resulting from the significant undermoderation in the upper part of the BWR core. A number of design modifications were proposed for eliminating or reducing the drawbacks associated with undermoderation of the upper part of a BWR core. In U.S. Pat. No. 3,145,149, entitled "Boiling Nuclear Reactor and Fuel Element Therefor" and issued Aug. 18, 1964, Imhoff suggested fuel rod designs in which the average quantity of fuel per unit length of fuel rod decreases towards the top of the core. Although this design reduces the degree of undermoderation in a BWR, the design significantly complicates the fuel rod fabrication. Furthermore, this design raises safety concerns. For example, fuel from upper pellets may disintegrate, fall through the central gap, and accumulate in a lower part of the fuel rod. This might lead to fuel and/or cladding meltdown which is a very undesirable accident. Gylfe, in U.S. Pat. No. 3,145,150, entitled "Fuel-Moderator Element for a Nuclear Reactor and Method of Making" and issued Aug. 18, 1964, proposed to use fuel rods having a double wall cladding. The inner clad was a tube made of a hydride material, e.g., zirconium hydride, that was filled with fuel pellets and cladded on the outside, with a thin layer of stainless steel or another material. As the number of hydrogen atoms per unit volume of zirconium hydride is comparable to hydrogen density in liquid water at room temperature, the zirconium hydride is an effective moderator material. Thus, the zirconium hydride as the primary fuel rod cladding material reduces the level of undermoderation in the upper part of a BWR core. A drawback of this scheme is that if the zirconium hydride cladding is made thick enough so as to provide significant moderation, the resistance of the zirconium hydride cladding to heat transfer significantly increases the fuel temperature. This may limit the power which can be extracted from a fuel rod and, hence, from a core of a given size. Weitzberg, in U.S. Pat. No. 4,591,479, entitled "Boiling Water Reactor Fuel Bundle," and issued on May 27, 1986, proposed to improve the moderation at the upper part of BWR cores by replacing fuel pellets at the upper part of a certain fraction of the fuel rods by pellets made of a solid moderator, such as zirconium hydride. Unfortunately, replacement of fuel pellets with zirconium hydride pellets reduces the amount of fuel and the total length of fuel rods in the core which in turn cancels most, if not all, of the improvement due to the increased moderation. Another drawback is that the zirconium of the zirconium hydride absorb a significant fraction of the moderated neutrons. Uchikawa et al., in U.S. Pat. No. 4,652,427 entitled "Fuel Assembly," and issued on Mar. 24, 1987, proposed to incorporate in a BWR fuel assembly a number of rods which contain burnable poisons mixed with a solid moderating material. A drawback is that this assembly does not significantly improve the undermoderation in the boiling part of the core. Another drawback is that it limits the amount of fuel and the total length of fuel rods which can be loaded into the BWR core. Doshi, in U.S. Pat. No. 4,664,882, entitled "Segmented Fuel and Moderator Rod," and issued on May 12, 1987, proposed to improve the moderation in the upper part of a BWR core by replacing one or more conventional fuel rods with segmented rods which contain fuel pellets in their lower part and water in their higher part. A drawback of this invention is that it reduces the amount of fuel and the total length of fuel rods in the core. Taleyarkhan, in U.S. Pat. No. 4,818,478, entitled "BWR Fuel Assembly Mini-Bundle Having Interior Fuel Rods of Reduced Diameter," and issued on Apr. 4, 1989, proposed to improve the moderation across the BWR fuel assembly by dividing an 8.times.8 fuel rod lattice into four 4.times.4 bundles that are separated by a cross shaped water gap in between. Moreover, the four inner fuel rods of each bundle are to have a smaller diameter, so as to provide more volume for water. A drawback of Taleyarkhan's invention is that it is more complicated than present BWR fuel assembly designs. Another drawback of this invention is that it does not significantly reduce the large variation in the degree of moderation along the fuel assembly. Typically, modern BWRs have a thermal power rating of 3000 to 4000 Megawatts, a fuel rod length of about four meters and a fuel rod outer diameter of about 1.25 centimeters (cm). The fuel, in the form of cylindrical pellets, is enclosed within a zircaloy, i.e., a zirconium alloy tube (also referred to as the cladding), nearly 0.9 millimeters (mm) in thickness. The fuel used by all BWRs is uranium oxide (UO.sub.2). In fact, with very few exceptions, oxide fuel is used in all the commercial power reactors operating around the world, including in pressurized water reactors (PWR), heavy water reactors (HWR) and even in liquid metal cooled reactors (LMR). The exceptions are a small number of gas cooled reactors (GCR) which use a metallic uranium alloy for their fuel. Metallic uranium alloy is also being considered for LMR under development in the USA. High temperature gas cooled reactors (HTGR) under development are designed to use uranium carbide and, possibly, also uranium oxide fuel. With one exception, the fuel for reactors used for research rather than for power production is a metallic uranium alloy, uranium oxide or uranium silicide. The exception is the so called TRIGA reactor which uses a hydride of a uranium-zirconium alloy for its fuel. Typically, the TRIGA reactor fuel rods are about 30 cm in length and about 3.5 cm in outer diameter and use 0.5 mm thick stainless steel cladding. The uranium-zirconium hydride composition used for the TRIGA fuel has, typically, 1.6 hydrogen atoms per zirconium atom, denoted as U-ZrH.sub.1.6. Details about the TRIGA fuel fabrication, properties and performance can be found in many publications, such as in the General Atomics report GA-A16029 by M. T. Simnad entitled "The U-ZrHx Alloy: Its Properties and Use in TRIGA Fuel" (Aug. 1980). The hydride fuel was selected for the TRIGA reactor primarily for its large negative effect on reactivity as the fuel temperature rises. This large temperature coefficient of reactivity enables the TRIGA reactor to generate power pulses, for the purpose of conducting various kinds of experiments. The power pulsing capability is one of the unique features of TRIGA reactors. Compared with a BWR core, a TRIGA core is small, has a low average power output and operates at low temperatures. Hence, a TRIGA core and the conditions under which a TRIGA core is operated are different from those of a BWR core. The small size of the TRIGA core means that larger fraction of fission neutrons leak from the core than in a BWR core. Thus, TRIGA reactor fuel needs a larger enrichment of uranium-235 to maintain the chain reaction. Traditionally, TRIGA fuel contained nearly 10 wt. % uranium that was enriched to more than 70% in U-235. Following the adoption of policies by the U.S. Government to limit the export of fuels to those enriched to less than 20% in U-235, General Atomic of San Diego, Calif. developed TRIGA fuel containing up to 45 wt. % uranium. For TRIGA and any other application of this fuel type, the uranium enrichment was 20% or very close to 20%. Whereas the former type of TRIGA fuel belongs to the category of Highly Enriched Uranium (HEU) fuel, the latter and present type of TRIGA fuel belongs to the category of Medium Enriched Uranium (MEU) fuel. Commercial light water reactors fuel use a Low Enriched Uranium (LEU) fuel. The uranium-zirconium hydride in the composition of this enriched fuel is reported by General Atomics to be stable and operational at temperatures up to 700.degree. C. (See above cited General Atomic reference). In fact, uranium-zirconium hydride fuel rods were successfully operated with linear-heat-rates comparable to the maximum linear-heat-rate a typical BWR fuel is designed to operate. The TRIGA core, in addition to its high enrichment and small size compared to a BWR core, is more symmetric than a BWR core. In particular, the partial water voiding associated with boiling is not encountered in a TRIGA core. Consequently, the upper part of the TRIGA core is as well moderated as its lower (or any other) part, and its power distribution is more symmetric than in a BWR core. Further, the TRIGA core is not subjected to the safety issues encountered in BWR cores due to the voiding of the upper part of these cores. Typically, neither the TRIGA reactor, nor any other type of research or power reactor, uses a combination of hydride and non-hydride fuel materials for their fuel. Moreover, due to the low weight percent (of about 10%) and the high enrichment (above about 70%) of the HEU in the uranium-zirconium hydride fuel developed for the original TRIGA reactors and used for many years in many TRIGA reactors around the world, the uranium-zirconium hydride has been considered as inadequate as a fuel for power reactors such as BWRs. Even the MEU fuel developed in the mid-seventies for TRIGA reactors is expensive and commonly considered as inadequate as a fuel for power reactors such as BWRs. Thus, for example, Glasstone and Sesonske, in their well known "Nuclear Reactor Engineering" text and reference book (Van Nostrand Reinhold Co., Third Edition, 1981), do not even refer to uranium-zirconium hydride, or to any other hydride material as a candidate fuel material for nuclear power reactors. The types of fuel materials they refer to are metallic, oxide, carbide and nitride. SUMMARY OF THE INVENTION According to the principles of this invention, novel fuel rods and a novel assembly of these fuel rods are used in a boiling water reactor (BWR) to improve the reactor performance in comparison to conventional BWRs that are in use for electric power generation. In the prior art BWRs, the fuel was made of uranium oxide (UO.sub.2) and was referred to as an oxide fuel. According to the principles of this invention, in regions where undermoderation occurs, a hydride fuel is substituted for the oxide fuel. Thus, one embodiment of the novel fuel rod of this invention has a first length fueled with an oxide fuel and a second length fueled with a hydride fuel. Whereas the oxide fuel has but a very small contribution to the slowing-down of the fission neutrons, the hydride fuel contains a large enough quantity of hydrogen and makes a good moderator in addition to being a fuel material. Thus, by using the hydride fuel instead of oxide fuel in undermoderated regions of the fuel assembly, the neutron moderation is improved without reducing the overall length of fuel in the assembly. Similarly, by replacing water rods, or rods containing another non-fuel moderator (such as zirconium hydride) with hydride fuel, the total length of fuel in the assembly is increased without significantly reducing the moderation ability of the fuel assembly. Relative to prior-art designs, the fuel assemblies of this invention increase the total power and/or total amount of energy which can be extracted from a BWR core of a given size. In particular, the fuel rods preferably include hydride fuel pellets at selected axial and radial positions within the core so as to improve axial and radial power profiles. Moreover, by placing hydride fuel in the upper part of the BWR core, the neutron moderation ability of this core region increases and the leakage of high energy neutrons is reduced. Therefore, the fission density in this core region increases, making additional contribution to the flattening of the power distribution. The use of hydride fuel permits elimination of water rods, part of the burnable poisons and reduction of the use of control rods in the prior art BWRs to achieve more uniform power distribution and burnup over the core during the irradiation cycle. It also improves safety and simplifies control of BWRs. According to the principles of this invention, the basic geometry of the BWR core including the fuel rods and fuel assemblies is not changed. Rather, the fuel pellets used in the fuel rods and the axial and radial composition of the fuel within a fuel assembly are modified to provide enhanced performance. While axial and radial dimensions are referred to herein with respect to a BWR core, these dimensional references are illustrative only of the principles of this invention and are not intended to limit the invention to the particular dimensions described. More generally, the axial dimension is a dimension in a first direction and the radial dimension is a dimension in a second direction where the second direction is orthogonal to the first direction. The coolant flow through the BWR core is in the first direction. In one embodiment of this invention, the oxide fuel pellets are located in that part of the fuel rod surrounded by non-boiling coolant and by boiling coolant with a steam volume fraction, also referred to as void fraction, of less than about 40%. The hydride fuel pellets are located in the region surrounded by boiling coolant that has a void fraction from about 40% up to the void fraction at the coolant exit from the core, i.e., the high void fraction part of the core. However, the starting point for the hydride fuel pellets may be any place in the range of from the onset of boiling up to a 70% void fraction. The actual position is selected so as to maintain as near as possible a uniform power distribution in the direction of coolant flow. Thus, since the fuel rod has a total fueled length in a first direction, a first fueled length is occupied by the oxide fuel pellets. A second fueled length is occupied by the hydride fuel pellets. Henceforth, a fuel rod containing both oxide fuel pellets and hydride fuel pellets is referred to as a mixed hydride-oxide fuel rod. A fuel rod containing oxide fuel pellets only is referred to as an all-oxide fuel rod. A fuel rod containing hydride fuel pellets only is referred to as an all-hydride fuel rod. There are various possibilities for arranging mixed hydride-oxide fuel rods and all-hydride fuel rods within a fuel assembly. In one embodiment, a number of the innermost fuel rods are all-hydride fuel rods, whereas the rest of the fuel rods are mixed hydride-oxide fuel rods. The all-hydride fuel rods can be arranged in different ways in the inner part of the fuel assembly. In one arrangement they are adjacent to each other, forming a region of all-hydride fuel rods at the center region of the assembly. In another arrangement they are intermixed with mixed hydride-oxide fuel rods in the inner part, i.e., center region, of the assembly. In a second embodiment, a first plurality of fuel rods in a fuel assembly are mixed hydride-oxide fuel rods, a second plurality of fuel rods in the fuel assembly are all-hydride fuel rods, and a third plurality of fuel rods in the fuel assembly are all-oxide fuel rods. Many other embodiments are possible using the novel fuel rods of this invention. One embodiment uses all-hydride fuel rods located at the inner region of the fuel assembly with the rest of the fuel rods being all-oxide fuel rods. Another embodiment uses mixed hydride-oxide fuel rods throughout the assembly. Yet another embodiment uses only mixed hydride-oxide fuel rods and all-oxide fuel rods. The novel fuel rods and fuel assemblies of this invention reduce the cost of generating electricity in BWRs. The BWR economics is improved by reducing the fraction of the fission neutrons that are absorbed in burnable poisons and in control rods and that leak-out through, primarily, the undermoderated portions of the core, thus improving BWR fuel utilization. The BWR economics is improved also by increasing the amount of energy which is extracted from a given fuel loading. This increase in energy generation and, hence, also in the fuel residence time in the core, is due to the increase in the overall fuel loading in the core and to the flatter power density distribution across the core. In addition, the safety of BWRs is improved by reducing the maximum power density these reactors need to operate at if they are required to deliver a given power output. The present fuel assemblies further improve the safety of BWRs by increasing the shutdown reactivity margin as well as by maintaining the effectiveness of the control rods more constant throughout the fuel cycle. Yet another improvement of the present fuel assemblies is a reduction of the fast neutron leakage from the upper part of the BWR core and, consequently, a reduction in the neutron induced damage rate to the structural components located near by the core. |
044302924 | abstract | In a system for disposing radioactive gaseous wastes in a nuclear power plant including a steam turbine and a main condenser connected thereto, a recombining unit is connected to the main condenser and includes a preheater for heating the radioactive gaseous wastes fed from the main condenser to a predetermined temperature and a recombiner for recombining into water vapor oxygen and hydrogen contained in the radioactive gaseous wastes passing through the preheater. A condenser is connected to the recombining unit for cooling the recombined water vapor contained in the radioactive gaseous wastes into condensed water and a hold-up device connected to the condenser for adsorbing and holding up the radioactive gaseous wastes with an adsorbing agent. The waste gas is discharged from a stack connected to the holdup device into atmosphere. |
053751523 | claims | 1. A method for preventing or reducing the formation of material contaminated with Co-60 on the surfaces of a circuit carrying cooling water in a nuclear reactor, comprising the steps of: adding at least one iron compound to said cooling water in an amount sufficient to maintain an iron concentration in said cooling water in a range from 50 to 200 ppb. adding at least one iron compound to said cooling water in an amount sufficient to maintain an iron concentration in said cooling water in a range from 50 to 200 ppb during shutdown of said reactor; and adding at least one iron compound to said cooling water in an amount sufficient to maintain an iron concentration in said cooling water in a range from 50 to 100 ppb during operation of said reactor after said shutdown. shutting down said reactor; removing said Co-60 contaminated material formation from said surfaces of said cooling water circuit; adding at least one iron compound to said cooling water to provide an iron concentration sufficient to scavenge cobalt from said cooling water; and after the cobalt has been scavenged, injecting oxygen into said cooling water to provide a dissolved oxygen concentration sufficient to form an oxide film on said surfaces of said cooling water circuit, said oxide film being substantially free of Co-60 isotope. 2. The method as defined in claim 1, wherein said iron compound is Fe(OH).sub.3. 3. The method as defined in claim 1, wherein said iron compound is Fe.sub.2 O.sub.3. 4. The method as defined in claim 1, wherein said iron compound is Fe.sub.3 O.sub.4. 5. The method as defined in claim 1, wherein said iron compound is ferrous oxalate. 6. The method as defined in claim 1, wherein said iron compound is ferric citrate. 7. The method as defined in claim 1, further comprising the step of maintaining a dissolved oxygen concentration in said cooling water in a range from 200 to 400 ppb. 8. The method as defined in claim 1, further comprising the step of heating said cooling water to a temperature of at least about 230.degree. C. and maintaining said temperature while the iron concentration in said cooling water is in said 50 to 200 ppb range. 9. The method as defined in claim 1, further comprising the step of treating said cooling water to attain a pH thereof in the range of about 7.5 to about 8.0 measured at a cooling water temperature of about 25.degree. C. prior to addition of said iron compound. 10. The method as defined in claim 9, wherein the pH and the iron concentration of said cooling water are maintained within said ranges for a period of at least 500 hours. 11. A method for preventing or reducing the formation of material contaminated with Co-60 on the surfaces of a circuit carrying cooling water in a nuclear reactor, comprising the steps of: 12. The method as defined in claim 11, further comprising the step of maintaining a dissolved oxygen concentration in said cooling water in a range from 200 to 400 ppb during said shutdown. 13. The method as defined in claim 11, further comprising the step of heating said cooling water to a temperature of at least about 230.degree. C. and maintaining said temperature while the iron concentration in said cooling water is in said 50 to 200 ppb range during said shutdown. 14. The method as defined in claim 11, further comprising the step of treating said cooling water to attain a pH thereof in the range of about 7.5 to about 8.0 measured at a cooling water temperature of about 25.degree. C. prior to addition of said iron compound during said shutdown. 15. The method as defined in claim 9, wherein the pH and the iron concentration of said cooling water during said shutdown are maintained within said ranges for a period of at least 500 hours. 16. A method for decontaminating the surfaces of a circuit carrying cooling water in a nuclear reactor, said surfaces being contaminated with a material formation containing Co-60 isotope, comprising the steps of: 17. The method as defined in claim 16, further comprising the step of adding at least one iron compound to said cooling water in an amount sufficient to maintain an iron concentration in said cooling water in a range from 50 to 100 ppb during operation of said reactor after said shutdown. 18. The method as defined in claim 16, further comprising the step of heating said cooling water to a temperature of at least about 230.degree. C. and maintaining said temperature while the iron concentration in said cooling water is 50 to 200 ppb. 19. The method as defined in claim 18, further comprising the step of treating said cooling water to attain a pH thereof in the range of about 7.5 to about 8.0 measured at a cooling water temperature of about 25.degree. C. prior to addition of said iron compound during said shutdown. 20. The method as defined in claim 18, wherein said dissolved oxygen concentration of said cooling water is 200 to 400 ppb. |
042016258 | claims | 1. A process for producing .sup.52 Mn by nuclear reaction, comprising the steps of: bombarding a vanadium-containing target with accelerated .sup.3 helium ions, and isolating the .sup.52 manganese thereby produced from the other target constituents after the bombardment by means of a chemical separation procedure. 2. A process as defined in claim 1 in which said vanadium-containing target is a metal foil of a substance selected from the group consisting of vanadium and vanadium alloys. 3. A process as defined in claim 1 in which a waiting period for the substantial decay of .sup.52 manganese is provided between the irridation bombardment of the target and the chemical isolation of the manganese produced. 4. A process as defined in any of the preceding claims in which in the step of bombarding the target, said accelerated .sup.3 helium ions are accelerated to an energy of about 14 MeV. 5. A process as defined in any of claims 1-3 in which said vanadium-containing target has a vanadium content consisting of .sup.51 V-enriched vanadium. 6. A process as defined in any of claims 1-3, in which said chemical separation procedure comprises extraction by a solution, in an organic liquid of a manganese-complexing agent. |
abstract | A process for controlling the dissolution of a metal in an acid bath is described. The metal may comprise aluminum and the acid bath may contain a metal catalyst that causes the metal to dissolve. In order to control the rate of dissolution and/or the amount of gas evolved during the process, an iron source is added to the bath. In one embodiment, the process can be used to dissolve aluminum contained in spent fuel assemblies for recovering a nuclear fuel, such as uranium. |
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050135198 | summary | BACKGROUND OF THE INVENTION This invention relates to a fast breeder reactor system capable of supplying a thermal/electrical output corresponding to the required scale. More particularly, the invention relates to an autonomous, decentralized fast breeder reactor system suited also to an underground site. Owing to the basic characterizing feature of a fast breeder reactor, namely the fact that fast neutrons are used in production of fission energy and in the breeding of fuel, the breeder has a dense core. Accordingly, in a larger scale power reactor, it is necessary to take a change in core reactivity into consideration in order to accumulate a larger volume of fuel within the core. In this connection, an inherent safety mechanism peculiar to fast breeder reactors is required. For example, since reactivity is controlled so as to be negative by a Doppler change or the like, core safety is maintained at a high level. However, in order to assure the reliability of these mechanisms, design limitations are imposed upon the core structure and control, etc. For example, a flat core structure, a structure which allows core expansion, or control for leveling core neutron flux is necessary. These limitations will not be solved merely by reducing the size of the core. There is a need to realize a universally applicable optimum nuclear reactor system in which rationalization of design is achieved based on a harmonious balance between safety and economy. SUMMARY OF THE INVENTION An object of the present invention is to provide an autonomous, decentralized fast breeder reactor system which operates in an autonomous manner while exhibiting a high degree of safety and reliability, and wherein scale of power can be varied easily in conformity with demand and application. In order to attain the foregoing object, the present invention provides an autonomous, decentralized fast breeder reactor system characterized in that a plurality of small-size nuclear reactor subsystems each having a small-scale faster breeder reactor core and a plurality of steam generator subsystems are disposed in a single main vessel of a nuclear reactor, and a heat transfer is made to take place between each nuclear reactor subsystem and each steam generator subsystem by a coolant circulating naturally through the interior of the main vessel of the nuclear reactor and undergoing a heat exchange with a coolant circulating naturally through the interior of each nuclear reactor subsystem. Other features and advantages of the present invention will be apparent from the following description taken in conjunction with the accompanying drawings, in which like reference characters designate the same or similar parts throughout the figures thereof. |
051868873 | summary | BACKGROUND OF THE INVENTION The present invention relates to an apparatus for inspecting the peripheral surfaces of cylindrical nuclear fuel pellets for defects while rotating the pellets on their own axes. In the manufacture of nuclear fuel pellets of uranium dioxide, very severe quality control is required. In particular, those defects which cannot be found by visual inspection must be found by non-destructive inspection. However, with the conventional inspection apparatuses for peripheral surfaces of nuclear fuel pellets, much time has been required for inspecting and sorting the pellets. SUMMARY OF THE INVENTION It is therefore the object of the present invention to provide an inspection apparatus for peripheral surfaces of nuclear fuel pellets by which the nuclear fuel pellets can be inspected automatically and sorted quickly and correctly. According to the invention, there is provided an apparatus for inspecting peripheral surfaces of nuclear fuel pellets, comprising: a handling unit for holding a prescribed number of nuclear fuel pellets in a line and rotating the same on axes thereof; an image pick-up device disposed adjacent to the handling unit for picking up image data as to the peripheral surfaces of the nuclear fuel pellets; a judging device operably connected to the image pick-up device for analyzing the image data outputted from the image pick-up device to output judging signals; and a sorting unit operably connected to the judging device for separating defective pellets from non-defective pellets based on the judging signals, the sorting unit including a plurality of sorting members disposed adjacent to the handling unit so as to correspond to the nuclear fuel pellets, respectively, and operating means operably connected to the judging device and the sorting members for operating the sorting members based on the judging signals. |
claims | 1. A system for detecting at least one contamination species in a vacuum, the system including:at least one cathode having a surface comprising a surface material, the at least one cathode configured to emit electrons into the vacuum; andat least one detector configured to detect the electrons emitted by the at least one cathode,wherein the cathode is configured to emit a first current of electrons when a cathode surface is substantially not contaminated by the at least one contamination species, andwherein the cathode is configured to emit a second current of electrons when the cathode surface is contaminated by the at least one contamination species based on an interaction between the at least one contamination species and the surface material. 2. The system according to claim 1, wherein the surface material comprises Lanthanum Hexaboride. 3. The system according to claim 1, further comprising:at least two cathodes having different surface areas. 4. The system according to claim 1, wherein the at least one detector comprises at least one anode configured to receive the electrons that are emitted by the at least one cathode. 5. The system according to claim 1, further comprising:an alarm generator configured to generate an alarm signal if the second current of electrons, detected by the at least one detector, is equal to or less than a threshold value. 6. The system according to claim 5, wherein the threshold value is substantially 0. 7. The system according to claim 1, wherein the second current of electrons is approximately equal to, or less than, half the first current of electrons. 8. The system according to claim 1, wherein the at least one contamination species comprises one or more hydrocarbon chains. 9. The system according to claim 1, wherein the surface material comprises at least one carbon nanotube. 10. A lithographic apparatus comprising at least one system according to claim 1. 11. The lithographic apparatus according to claim 10, further comprising:an illumination system configured to provide a radiation beam;a patterning device configured to impart the radiation beam with a pattern in its cross-section;a substrate holder configured to hold a substrate; anda projection system configured to project the patterned radiation beam onto a target portion of the substrate. 12. The system according to claim 1, wherein the surface material comprises Cesium Hexaboride. 13. A contamination detection system, comprising:a chamber;a cathode situated in the chamber, the cathode comprising a surface coated in a surface material being configured to emit electrons as a first current of electrons, the electrons comprising a first current of electrons when the cathode surface is substantially free of contamination from at least one contamination species, and the electrons comprising a second current of electrons when the cathode surface is contaminated by the at least one contamination species based on an interaction between the at least one contamination species and the surface material;a detector configured to detect the electrons emitted from the cathode; anda controller configured to generate an alarm signal when the detector detects the second current of electrons. 14. The system of claim 13, further comprising:a plurality of cathodes situated in the chamber; anda plurality of detectors corresponding to the plurality of cathodes. 15. The system of claim 14, wherein:a first cathode from among the plurality of cathodes comprises a first surface coated in a first surface material; anda second cathode from among the plurality of cathodes comprises a second surface coated in a second surface material. 16. The system of claim 15, wherein the first surface material comprises a different material than the second surface material. 17. The system of claim 13, wherein the surface material comprises Lanthanum Hexaboride. 18. The system of claim 13, wherein the surface material comprises Cesium Hexaboride. 19. The system of claim 13, wherein the second current is less than the first current. 20. The system of claim 13, further comprising:a current source coupled to the cathode and configured to generate an electric current to heat the cathode to remove the at least one contamination species after the detector has detected the second current of electrons. |
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046860797 | summary | BACKGROUND OF THE INVENTION This invention relates to a fuel assembly, and more particularly to a fuel assembly having a fuel spacer consisting of a large number of circular members. As a fuel assembly for use in a boiling water reactor, a fuel assembly is known which has a fuel spacer consisting of a large number of circular sleeves. This fuel assembly is disclosed in Japanese patent Laid-Open No. 65287/1984 (basic application: U.S. Patent application Ser. No. 410124 filed on Aug. 20, 1982), and is illustrated in FIG. 2 of the Japanese reference. The fuel spacer is produced by arranging a large number of circular sleeves, into which fuel rods are inserted, in grid form and coupling adjacent circular sleeves with one another by welding. A water rod is also inserted into the circular sleeve in the same way as the, fuel rods. The support of the fuel spacer by the water rod is shown in FIGS. 8A-8C and 9A-9B of the Japanese reference. Development of a fuel assembly, which includes a water rod having an increased outer diameter, has also been made. An example of such fuel assemblies is shown in FIG. 1 of Japanese patent Laid-Open No. 65792/1984. The outer diameter of the water rod in this fuel assembly is about twice the outer diameter of the fuel rod. SUMMARY OF THE INVENTION It is an object of the present invention to provide a fuel assembly with low pressure loss, which can increase a flow path area inside the water rod. It is another object of the present invention to provide a fuel assembly which can reduce the danger of contact between the fuel rods and the water rod. One of the characterizing features of the present invention resides in that the fuel spacer includes bridge members, whose both ends are fitted to two adjacent circular members among those circular members which are arranged close to a water rod having an outer diameter greater than that of fuel rods, in such a manner as to encompass the water rod, and projections disposed on the side surfaces of the water rod support the bridge members of the fuel spacer. Another characterizing feature of the present invention is that both ends of each bridge member exist between the water rod and the circular members to which the bridge member is fitted. |
claims | 1. A method of generating a safety demand signal for a nuclear power plant, the method comprisingreceiving a plurality of first sensor signals and a plurality of second sensor signals at a first division;receiving a plurality of third sensor signals and a plurality of fourth sensor signals at a second division;generating first and second data signals based on the first and second sensor signals respectively;generating third and fourth data signals based on the third and fourth sensor signals respectively;sending the first and second data signals from the first division to the second division;sending the third and fourth data signals from the second division to the first division;determining that one of the first or second sensor signals is erroneous;determining that one of the third or fourth sensor signals is erroneous;entering a limiting condition of operation to accommodate a single failure criterion non-compliant condition;changing voting logic in the first division based on the determining that either one of the first or second sensor signals is erroneous;generating a first intermediate safety demand based on the changed voting logic in the first division and at least two of the first, second, third, and fourth data signals or one of the first or second data signals;generating a second intermediate safety demand based on the changed voting logic in the first division and at least two of the first, second, third, and fourth data signals or one of the first or second data signals;generating a first final safety demand based on the first intermediate safety demand and the second intermediate safety demand; andending the limiting condition of operation when single failure criterion compliance is restored,whereinthe first final safety demand indicates that a reactor trip or engineered safety features actuation is necessary only if both the first intermediate safety demand and the second intermediate safety demand indicate that a reactor trip or engineered safety feature actuation is necessary, andthe plurality of first sensor signals, the plurality of second sensor signals, the plurality of third sensor signals, and the plurality of fourth sensor signals all measure the same system parameter. 2. The method of claim 1, whereinthe changed voting logic is two-out-of four voting with respect to the first, second, third, and fourth data signals, or one-out-of-two voting with respect to the first or second data signals when there is a signal failure that puts the plant in a single failure criterion non-compliant configuration. 3. The method of claim 1, whereinthe changed voting logic is A-out-of-B voting with respect to the first, second, third, and fourth data signals, or C-out-of-D voting with respect to the first or second data signals,B is a total number of available data signals from among the first, second, third, and fourth calculation result signals, andA is an integer equal to or lower than B,D is a total number of available data signals from among the signals originating within the same division, andC is an integer equal to or lower than D. 4. The method of claim 1, further comprisingchanging voting logic in the second division based on the determining that either one of the third or fourth sensor signals is erroneous;generating a third intermediate safety demand based on the changed voting logic in the second division and at least two of the first, second, third, and fourth data signals or one of the third or fourth data signals;generating a fourth intermediate safety demand based on the changed voting logic in the second division and at least two of the first, second, third, and fourth data signals or one of the third or fourth data signals; andgenerating a second final safety demand based on the third intermediate safety demand and the fourth intermediate safety demand; andgenerating a reactor trip or engineered safety features actuation by using one-out-of-two voting based on the first and second final safety demands from the first and second divisions, respectively. 5. The method of claim 1, whereinthe generating of the first intermediate safety demand is performed by using one-out-of-two voting between a first result of one-out-of-one voting based on one of the first and second data signals and a second result based on two-out-of-three voting with respect to three of the first, second, third, and fourth data signals. 6. The method of claim 1, whereinthe generating of the first intermediate safety demand is performed by using one-out-of-two voting between a first result of one-out-of-one voting based on one of the first and second data signals and a second result of two-out-of-two voting with respect to two of the first, second, third, and fourth data signals. 7. The method of claim 1, whereinthe generating of the first intermediate safety demand is performed in the first division,the generating of the second intermediate safety demand is performed in the first division,the first intermediate safety demand and the second intermediate safety demand are combined in subsequent two-out-of-two logic in the first division to generate a first final reactor trip or engineered safety features actuation demand, andthe first intermediate safety demand and second intermediate safety demand use logic that resides in different reactor protection processor memory locations. 8. The method of claim 1, further comprisingdelaying the first intermediate safety demand by a first delay amount prior to generating the first final reactor trip or engineered safety features actuation demand; anddelaying the second intermediate safety demand by a second delay amount prior to generating the first final reactor trip or engineered safety features actuation demand. 9. The method of claim 1, further comprising:energizing an energize-to-actuate shunt trip coil to activate a reactor trip breaker based at least in part on the first final reactor trip or engineered safety features actuation demand. 10. The method of claim 1, further comprising:energizing the shunt trip coil and de-energizing an undervoltage coil to actuate a reactor trip breaker in the first division based on an energize-to-actuate final safety demand from the safety division. 11. The method of claim 10, further comprising:providing a first watch-dog-timer in the first division, which de-energizes when a first reactor protection processor failure or first loss of power is detected within the first division,providing a second watch-dog-timer in the second division, which de-energizes when a second reactor protection processor failure or a second loss of power is detected within the second division,configuring the first and second watch-dog-timers so that they de-energize first and second undervoltage coils in first and second reactor trip breakers when both the first and second watch-dog-timers detect a reactor protection processor failure or loss of power. 12. The method of claim 1, further comprising:providing a first watch-dog-timer output at a first division reactor protection processor, the first watch-dog-timer being associated with the first division and indicating whether or not a reactor trip is required;de-energizing a de-energize-to-activate undervoltage coil to activate a reactor trip breaker based at least in part on an output of the first watch-dog-timer output. 13. The method of claim 1, further comprising:changing voting logic in the second division based on the determining that either one of the third or fourth sensor signals is erroneous;generating a third intermediate safety demand based on the changed voting logic in the second division and at least two of the first, second, third, and fourth data signals or one of the third or fourth data signals;generating a fourth intermediate safety demand based on the changed voting logic in the second division and at least two of the first, second, third, and fourth data signals or one of the third or fourth data signals;generating a second final safety demand based on the third intermediate safety demand and the fourth intermediate safety demand,wherein the second final safety demand indicates that a reactor trip or engineered safety features actuation is necessary only if both the third intermediate safety demand and the fourth intermediate safety demand indicate that a reactor trip or engineered safety features actuation is necessary. 14. The method of claim 1, whereinthe method does not employ any other divisions than the first and second divisions in determining whether a reactor trip is necessary. 15. The method of claim 1, whereinthe first division includes a first reactor trip breaker and a second reactor trip breaker, andthe first reactor trip breaker is in parallel with the second reactor trip breaker,further comprisinggenerating a first reactor trip initiation signal by the first division when both the first and second reactor trip breakers are activated. 16. The method of claim 15, whereinthe second division includes a third reactor trip breaker and a fourth reactor trip breaker, andthe third reactor trip breaker is in parallel with the fourth reactor trip breaker,further comprisinggenerating a second reactor trip initiation signal by the second division when both the third and fourth reactor trip breakers are activated. 17. The method of claim 16, further comprisinginitiating a reactor trip or engineered safety feature actuation when either the first reactor trip initiation signal is generated or the second reactor trip initiation signal is generated. |
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047553286 | abstract | The invention relates to a process for decontaminating and adjusting the pH of uraniferous solutions to render them compatible with the natural environment into which they may be discharged. This process is characterized in that the solutions to be treated having a natural pH from about 2.5 to about 6.5 and containing from about 1 to about 100 mg/l of uranium, are supplemented with an aluminum salt, such as sodium aluminate, in a sufficient amount for the final pH to be from about 5.5 to about 8.5 and for there to be precipitation, coagulation and adsorption of about 90% of the uranium initially contained in the solution and for the uranium content remaining in the final solution obtained to be equal to or less than about 1.8 mg/l. |
abstract | A liquid metal ion gun 3 includes a liquid metal ion source 31 and a beam limiting aperture 33. The liquid metal ion source 31 includes a reservoir 36 and an emitter 35. The reservoir 36 is made of tungsten (W) and holds liquid metal gallium (Ga). The emitter 35 is made of W. The beam limiting aperture 33 is formed with a liquid metal member 44 made of Ga placed on a base 46 made of W, has an opening 41 that enables an ion beam 2 extracted from the liquid metal ion source 31 to pass therethrough, and limits the diameter of the ion beam 2. The beam limiting aperture 33 has a groove structure 45 that causes the liquid metal 44 to gather into a region located around the opening 41. The lifetime of the beam limiting aperture can be increased, and an emission can be maintained stable for a long time period and reproducibly restored to a stable state. |
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044434030 | description | DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows the structure 1 of a nuclear reactor bearing a vessel 2 inside which fuel assemblies 3 are disposed, constituting the core of the nuclear reactor. A transfer machine 5, which can be moved in rotation by means of rotating plugs 7 and 8 so as to occupy any position above the lattice of fuel assemblies, assures transfer of the assemblies 3 first to position 3a by means of telescopic movements of the transfer machine, then to position 3b by rotation of the transfer machine and then lowering of the assembly vertically by telescopic movements. In position 3b, the assembly can be tilted into a ramp termed the "primary ramp" which opens at its upper part into a lock chamber 11 inside which a ramp 12 also opens, so that the lock chamber communicates with a storage region 14. Isolation of the lock chamber with respect to the ramps 10 and 12 can be obtained by means of valves 16 and 17. A ramp 15 movable in rotation and solid with a rotating plate 18 disposed at the upper part of the lock chamber 11 can come successively into position 15a in communication with the primary ramp and position 15b in communication with the secondary ramp. An assembly 3 can be brought inside this movable ramp from the primary ramp or from the secondary ramp by means of a hoist 19 mounted above the lock chamber 11 which is termed the "loading and unloading lock chamber". Moving assemblies inside the ramps 10 and 12 and inside the lock chamber can be carried out the hoist 19 alone or by in combination with other hoists disposed at the entry to the primary and secondary ramps. At the exit from the ramp 12, the assembly can, after tilting into vertical position, be taken up by a transfer machine 20 which can move the assembly 3 in the vertical direction in both directions, allowing assembly 3 to be placed in any position inside the rotating storage drum 21 disposed inside the storage region 14. Conversely, new assemblies disposed in the storage region can be raised by the transfer machine 20 and disposed in the part of this region which allows them to tilt into the secondary ramp 12 so that these assemblies can be introduced into the reactor vessel by means of the loading or unloading lock chamber 11. The loading and unloading lock chamber is permanently disposed on the structure 1 of the nuclear reactor which supports the ramps 10 and 12 opening into the lock chamber. It is clear that, in this apparatus according to the prior art, the large lock chamber forms a complex and bulky unit covered by very thick casings 23 to protect the regions near the lock chamber from radiation. FIGS. 2 and 3 show a loading and unloading apparatus according to the invention in its positions for service on the reactor and uncoupled with respect to this reactor. Such an apparatus according to the invention has a closed, sealed, metal vessel 25 with two apertures 26 and 27 corresponding respectively to the primary and secondary ramps 28 and 29 of the reactor on which the removable lock chamber according to the invention is positioned. Valves 30 and 31 respectively enable these apertures 26 and 27 located on the vessel to be closed, while the valves 32 and 33 respectively allow the ends of the primary and secondary ramps to be closed, these ramps communicating with the interior of the reactor and the storage region. Connection flanges disposed between the valves 30 and 32 and 31 and 33 respectively allow the removable lock chamber constituted by the sealed vessel 25 to be fixed on the primary and secondary ramps of a nuclear reactor. The apparatus for connection and isolation between the lock chamber and the reactor ramps will be described in further detail with reference to FIGS. 7, 8, 9 and 10. The vessel 25 is covered by elements 34 assuring shielding from radiation of the outside of the lock chamber with respect to the inside. A movable ramp 35 is disposed inside the lock chamber on a rotating element 36 which allows this ramp to be placed in communication with either the primary ramp 28 or the secondary ramp 29. Above the rotating unit 36 a hoist 37 is disposed which allows the fuel assemblies to be moved inside the ramps either upwards or downwards. FIG. 3 shows that, when uncoupling between the removable lock chamber and the reactor has been carried out by removal of the assembly flange disposed between the isolating valves, the removable lock chamber can be raised by means of an overhead travelling crane, connector elements 38 being provided on the upper parts of the lock chamber. Where several nuclear reactors are disposed on one and the same site, it is possible to transport the removable lock chamber to another nuclear reactor where coupling with the primary and secondary ramps can be effected. The apertures in the structure of the nuclear reactor from which the lock chamber has just been removed are closed by radiation shielding plates 40. FIG. 4 shows a removable lock chamber identical to the lock chamber represented in FIGS. 2 and 3, but having radiation shielding elements constituted by two separate parts 41 and 42 which can be placed in position on the vessel 25 or conversely removed therefrom independently of each other. In practice, the lock chamber and its radiation shielding elements form a very heavy unit, weighing 400 to 500 tons, for example, and it can prove difficult to have available lifting means sufficient to lift the lock chamber unit above the reactor and move it into a new service position above another reactor. In the case of the apparatus represented in FIG. 4, before the lock chamber is raised above the reactor for transportation, the pieces 41 and 42 are lifted and transported to a place near the nuclear reactor on which the lock chamber is to be disposed, before lifting and transportation of the latter is carried out as represented in FIG. 4. FIGS. 5 and 6 show another embodiment of the vessel of the removable lock chamber, this vessel 45 being constituted by a cylinder with two apertures 46 and 47 at its lower part and a unit 48 for moving the movable ramp 50 as well as a hoist 49 at its upper part. As FIG. 6 shows, this unit can be disconnected from the primary and secondary ramps of the reactor and extracted from the protective structure 51 constituted by a very thick cylinder, made of shielding material disposed at a fixed station on the reactor and having a cross-section corresponding to that of the vessel. A complementary protective plate 52 is then disposed on the upper part of the cylinder 51 to isolate the interior of the reactor totally. FIGS. 7 and 8 show in greater detail the whole of the means for connecting and isolating the lock chamber and the secondary ramp of the reactor. These connecting and isolating means are positioned at the apertures of the vessel 55 allowing the latter to be placed in communication with the primary ramp or with the secondary ramp. FIGS. 7 and 8 show the connecting and isolating means which allow the vessel to be put in communication with the secondary ramp 56 solid with the structure 57 of the reactor. The ramp 56 is solid with a unit comprising an isolating valve 58 and a mounting flange 59 bearing guide pins 60. The vessel 55 bears on its lower part, by a bellows joint 61, an isolating valve 62 and a ramp section 63 fixed with respect to the vessel 55, apart from slight displacements or deflections of this ramp 63 with respect to the vessel 55, obtained by deformation of the bellows joint 61 positioned in the extension of the ramp 56. The movable ramp 64 can then come into position as an extension of the ramp 63 by tilting. FIG. 7 shows the ramp 64 in its position for communication with the secondary ramp 56 and FIG. 8 shows the movable ramp 64 in a position in which the movable ramp and the fixed ramp are no longer extensions of each other. The movable ramp is displaceable by tilting and rotation between the position represented in FIG. 7 and a position in which the movable ramp 64 forms an extension of a fixed ramp section (not represented) allowing connection with the primary ramp of the reactor. The vessel 55 is surrounded by elements 65 protecting against radiation, which elements, when the lock chamber is put in position on the reactor, form extensions of protective elements 66 solid with the structure 57 of the reactor, so as to constitute a jointly therewith protection over the whole periphery of the lock chamber. A flange 67 is also solid with the valve 62 at its lower end and comes into position opposite the flange 59 solid with the valve 58 disposed on the fixed support structure of the secondary ramp of the reactor. The flange 67 has bores allowing guide pins 60 to be placed in position when the lock chamber docks with the reactor. As the removable lock chamber according to the invention is to serve for several reactors, disposed on one and the same site, for example, the connection apparatus of this lock chamber must be adaptable to slight differences in dimensions which may exist between two reactors, affecting the inclination or position of the primary and secondary ramps. For this reason ramp 63 solid with the vessel 55 is fixed to the latter by a bellows joint 61 which allows slightly displacements or deflections. In this way, the guide pins 60 of the flange 59 can always be made to coincide with the bores in the flange 67. During positioning of the two connecting and isolating apparatuses corresponding to each of the primary and secondary ramps, securing of the lock chamber is effected by clamping together the flanges 59 and 67 and the corresponding flanges at the primary ramp. The flanges 59 and 67 have a certain number of bores 68 for this, in which screws are introduced which are secured by nuts. When the movable ramp 64 comes into position by tilting with respect to the ramp 63 to form an extension of this, by means of a stop arresting its travel at the required place, continuity of the ramps is assured between the movable ramp and the primary and secondary ramps of the reactor. A slight angular variation in the displacement of the movable ramp allows accommodation of slight deviations arising from differences between the inclinations or spacings of the ramps of one reactor in relation to another. FIGS. 9 and 10 show another embodiment of the apparatus for connection and isolation between the vessel 55 and the secondary ramp 56, the vessel 55 bearing at its aperture 69 a flange 70 bearing guide pins 71. The upper end of the ramp 56 bears a valve 72 itself extended by a ramp portion 73 while a flange 74 with bores corresponding to the guide pins 71 is fixed on the valve 72 by a bellows joint 75. In this way, differences in angle or spacing between the ramps and the apertures of the lock chamber can be accommodated when the lock chamber is moved from one reactor to another. In this case, the tilter 76 comes into position as an extension of the ramp portion 73 by tilting. An inner valve 77 allows the aperture 69 to be closed when the lock chamber is isolated with respect to the external environment, for example for transport purposes. The principal advantages of the apparatus according to the invention are that (a) it allows use of only one lock chamber containing the movable apparatuses for loading and unloading the assemblies per group of reactors, thereby allowing saving on the costs of constructing the nuclear reactor, (b) it allows the mechanisms associated with the lock chamber to be operated more frequently; and (c) it allows easy maintenance of the lock chamber while the reactor is operating. Also, between periods of use of the lock chamber on the nuclear reactor, a large space is left available above the vessel of the reactor in the building for the latter. The invention is not, however, limited to the embodiments described; it also includes all the variants thereof. Thus, the shape of the vessel constituting the lock chamber is not limited to the shapes described and represented, the way in which the lock chamber is secured and isolated can be different from the securing and isolating methods described, and protection against radiation can be obtained differently from the particular methods described. Finally, the loading and unloading apparatus according to the invention can be used for all fast neutron nuclear reactors using a lock chamber moving the fuel elements from a primary ramp communicating with the reactor vessel to a secondary ramp communication with a position for storing or loading fuel assemblies. |
claims | 1. A method of seismic retrofitting a concrete structure comprising:removing material from a portion of the concrete structure by irradiating the portion with a laser beam having a laser energy density while moving the laser beam laterally across the portion;removing material from the concrete structure to form a generally quadrilateral-shaped recess;forming multiple such generally quadrilateral-shaped recesses in the concrete structure;positioning a temporary mold in proximity to the portion of the concrete structure; andattaching a stabilization structure to the portion of the concrete structure by pouring concrete into the temporary mold such that at least a portion of the recesses are filled with concrete, whereby the stabilization structure provides structural support to the concrete structure. 2. The method of claim 1, wherein the portion of the concrete structure comprises a wall and removing material from the portion comprises boring a hole into the wall. 3. The method of claim 2, wherein boring the hole comprises moving the laser beam in a circular motion along a surface of the wall such that a substantially cylindrical hole is formed. 4. The method of claim 1, wherein the portion of the concrete structure comprises a wall and removing material from the portion comprises cutting a key into the wall. 5. The method of claim 4, wherein cutting the key comprises moving the laser beam in multiple cutting passes along a surface of the wall. 6. The method of claim 4, wherein the key has a generally rectangular shape. 7. The method of claim 1, wherein the portion of the concrete structure comprises a rebar embedded in the concrete structure, and removing material comprises removing concrete to expose a portion of the rebar. 8. The method of claim 7, wherein removing material further comprises detecting the rebar and avoiding substantially irradiating the rebar, thereby avoiding substantially damaging the rebar. 9. The method of claim 8, wherein detecting the rebar comprises locating the rebar using x-rays. 10. The method of claim 8, wherein detecting the rebar comprises using an electronic eye to detect light reflected from the rebar as material is being removed. 11. The method of claim 10, wherein avoiding substantially irradiating the rebar comprises moving the laser beam away from the rebar upon detecting light reflected from the rebar. 12. The method of claim 10, wherein avoiding substantially irradiating the rebar comprises reducing the laser energy density upon detecting light reflected from the rebar. 13. The method of claim 1, wherein the portion of the concrete structure comprises a column. 14. The method of claim 13, wherein comprises moving the laser beam in multiple cutting passes along a surface of the column. 15. The method of claim 1, wherein the concrete structure is occupied by equipment and people, the equipment and people having a noise tolerance level, a vibration tolerance level, and a particulate tolerance level, and removing material generates noise at a noise level less than the noise tolerance level, vibrations at a vibration level less than the vibration tolerance level, and particulates at a particulate level less than the particulate tolerance level. 16. The method of claim 15, wherein the concrete structure comprises a healthcare facility and the equipment and people comprise healthcare equipment, personnel, and patients. 17. A method of seismic retrofitting a concrete structure comprising:removing material from a portion of the concrete structure by irradiating the portion with a laser beam having a laser energy density;positioning a stabilization structure in proximity to the portion of the concrete structure; andattaching the stabilization structure to the portion of the concrete structure, whereby the stabilization structure provides structural support to the concrete structure wherein removing material from the portion of the concrete structure comprises coring a cylindrical plug using the laser beam and breaking off the cylindrical plug from the portion of the concrete structure. |
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description | This application is a continuation of U.S. application Ser. No. 12/730,226 filed on Mar. 23, 2010, which claims the benefit of U.S. Provisional Patent Application No. 61/162,374 filed on Mar. 23, 2009, the entire disclosure of each of which is incorporated by reference herein. This invention was made with government support under HG004764 and HG003578 awarded by National Institute of Health and DE-AC02-05CH11231 and DE-AC05-00ER22725 awarded by Department of Energy. The government has certain rights in the invention. The present invention relates to systems and methods for trapping a charged particle in a liquid environment. Further, the present invention relates to systems and methods for controlling, sensing, and identifying trapped charged particles. Nanoscale control of matter has led to enormous advances in many fields. In the biological and medical fields continued advances will allow for an unprecedented ability to examine and manipulate biological molecules and reactions. To achieve this, efficient methods for trapping, identifying, and sensing properties of biomolecules are needed. One biomedical application in particular, genome sequencing, is a prime example of an application amenable to such a nanoscale approach. Current methods of genome sequencing such as chain-termination gel electrophoresis are slow and costly. This, coupled with the fact that a human genome contains approximately 3 billion base pairs makes sequencing even a single human genome a monumental task. The possibility of direct genome sequencing using electronic measurements, wherein each base pair is identified as it basses by a nanoscale sensor, is potentially orders of magnitude faster and proportionally less costly then existing methods. These new techniques could enable sequencing of any individual's genome to prevent, diagnose, and treat diseases, potentially leading to a new genome-based medical practice. One method of directly sequencing DNA involves translocating a fragment of Single-Stranded DNA (ssDNA) through a nanogap or a nanopore. These nanopores confine the DNA and allow for measurement of its properties as it translocates through the nanopore. Differences in the structure of the different nucleotides give rise to measurable effects which can be detected. Several measurements can distinguish between different bases, allowing for sequencing the DNA as it passes through the nanopore. If an ionic current is flowing through a nanopore, it has been found that DNA translocating through the pore masks the ionic current in a way specific to the nucleotide instantaneously passing through the pore. Alternatively, a bias applied across the transverse direction of the nucleotide can measure the capacitance or conductance of that specific nucleotide. Repeatable measurements of the base specific signature of each nucleotide depends critically on its relative geometry during translocation. For example, it has been found that the variation in the transverse conductance due to the geometry of a base relative to an electrode can easily outweigh the differences between different types of nucleotides. Differences in the orientation and position of nucleotides relative to sensors must be minimized to make such a system feasible. Because an ssDNA is only about a nanometer wide, trapping methods that can achieve control on this scale are required. Further, DNA sequencing occurs in an aqueous or electrolytic environment, and an appropriate method of trapping the DNA must be compatible in such conditions. In a broader context, however, the general techniques of trapping and manipulating particles in liquid environment at a nanoscale resolution are important for a number of applications beyond DNA sequencing. Specifically, many molecules of interest become charged upon dissociation in an aqueous environment, and such a method could enable efficient trapping, sensing, identifying and sorting of these molecules. Over the last few decades, a variety of manipulation techniques have been developed to achieve trapping of particles in liquids. These methods include optical tweezers, acoustic tweezers, and magnetic tweezers. These methods, however, can require complicated setups that have a low potential for integration into compact and cost effective devices. Because of this, increasing use has been made of electrical forces for achieving manipulations of particles in liquids. Dielectrophoresis (DEP) forces arise from an object's polarizability. By applying a nonuniform electric field, it is possible to induce a dipole moment on an uncharged particle and create either an attractive or repulsive force. Using DEP it is possible to trap small particles in solutions. Indeed, the electrical trapping of objects in solution has so far been done primarily by DEP. DEP forces, however, are relatively weak, especially for smaller targets since the forces scale with the volume of the trapped object. Particles with diameters below 1 μm, for example, cannot be trapped by DEP as Brownian motion overwhelms the DEP forces. For this reason DEP based traps are not attractive for detection of very small biomolecules, such as ssDNA bases for direct sequencing. Electrophoresis, in contrast, makes use of the interaction of an object's fixed charge and an electric field. Electrophoresis depends upon the amount of charge rather than polarizability, and is a first order interaction with the electrical field. While useful for moving particles, the multipole fields are unsuitable for trapping applications. This is because a charged particle cannot be stably held in a multipole electrostatic field due to the saddle shape of the potential that results from Laplace's equation. While a charged particle may be confined in one dimension, it will necessarily be unconfined in another. Although this would seem to preclude electrophoretic traps, one can get around this problem by using a time-varying field. One such system is the anti-Brownian electrophoretic trap (ABEL) based on a feed-back mechanism. In this system the computer visually tracks the trajectory of a charged particle. Using this information, the computer calculates a feedback voltage which is applied to electrodes arranged around a trapping volume containing the target particle. The applied feedback voltage creates an force to counteract the particles motion and return it to the center of the trapping volume. This technique requires a visible target and stability depends upon a fast sampling rate. These limitations make the technique unsuitable for many applications. There remains a need in the art for systems and methods for controlling charged particles in liquids. Preferably such a system would utilize the strong electrophoresis force without requiring complicated setups or detection schemes. Such a system is desirable as it could enable efficient control of biomolecules for a variety of applications including DNA sequencing. In contrast to trapping techniques in a liquid environment, trapping charged particles in vacuum and gaseous environments using electromagnetism is a mature field. It is known that atomic ions and other charged particles can be confined by particular arrangements of electromagnetic fields in these environments. One such device is a Paul trap, which can be used to dynamically confine particles in vacuum or gas through spatially inhomogeneous and alternating radio frequency (RF) electrical fields. In this type of device a set of electrodes generates an alternating quadrupole potential which can provide confinement in two or three dimensions. While at any given moment the potential within the trap is an unstable saddle point, changing the orientation of this saddle point rapidly by providing an appropriate RF field can in fact create a dynamically stable trap. Paul traps are used in vacuum and gaseous environments today for a number of applications including analytical chemistry and aerosol research, and their version, a linear Paul trap, is an important component of Mass Spectrometry instruments. While Paul traps exhibit many properties which are attractive as a potential trapping method for charged particles in liquids, it has been the general consensus that such a device was incompatible with a liquid environment. Polarization of the liquid, thermal fluctuations due to Brownian motion, charge screening, and viscosity were all effects indicated that such a device was impossible. To date, no Paul traps have been demonstrated in a liquid environment. Accordingly, presently there is a need in the art for Paul traps capable of trapping charged particles in liquids. Additionally, there is a need for incorporating these novel Paul traps into systems for controlling, sensing, and identifying charged particles in a liquid environment. A system for trapping a charged particle in a trapping volume is disclosed. The system comprises at least three confining electrodes distributed around the trapping volume. Between these three confining electrodes is a liquid container adapted to hold a liquid carrying the charged particle. A power source electrically connecting the at least three confining electrodes is capable of applying a time-varying periodic voltage bias between the at least three confinging electrodes for creation of a time-varying periodic multipole electric potential in the trapping volume. The multipole electric potential is at least a quadrupole electric potential. The trapping volume may be microscopic or nanoscopic. The liquid container can be fluidly connected to a microfluidic channel for supplying liquid to the liquid container. The liquid container is adapted to hold a liquid solution comprising an electrolyte. In some embodiments the three confining electrodes are N confining electrodes where N is an even whole number of four or greater. These N confining electrodes are positioned around the trapping volume so that the multipole electric potential in the trapping volume is orthogonal to a longitudinal axis to the trapping volume. These N confining electrodes may be coplanar in a plane orthogonal to the longitudinal axis of the trapping volume. Two electrodes are arranged along the longitudinal axis of the trapping volume so that the trapping volume is between them. These two electrodes are electrically connected to a power source capable of applying a voltage bias for controlling the movement of the charged particle along the longitudinal axis of the trapping volume. In another embodiment the system for trapping a charged particle comprises two longitudinally confining electrodes along a longitudinal axis. A transversely confining electrode encircles the region between the two longitudinally confining electrodes transversely to the longitudinal axis. A liquid container between the two longitudinally confining electrodes and the transversely confining electrode is adapted to hold a liquid carrying the charged particle. A power source is electrically connected to the longitudinally confining electrodes and the transversely confining electrode and is capable of applying a time-varying periodic voltage bias to create a time-varying quadrupole electric potential in the trapping volume. A method of trapping a charged particle in a trapping volume is also disclosed. The method comprises: providing a liquid containing the charged particle; positioning the liquid containing the charged particle between at least three confining electrodes; applying a time-varying periodic voltage bias to the at least three confining electrodes distributed around the trapping volume; generating a multipole electric potential in the trapping volume, wherein the multipole electric potential is at least a quadrupole; and trapping the charged particle within the trapping volume. In some embodiments of the method the at least three confining electrodes are N confining electrodes positioned around the trapping volume so that the multipole electric potential in the trapping volume is orthogonal to a longitudinal axis of the trapping volume, wherein N is an even whole number of four or greater. The N confining electrodes may also be coplanar in a plane orthogonal to the longitudinal axis of the trapping volume. The present invention advantageously provides the ability to trap charged particles. The term “trapping” and its variations indicate that movement of the charged particle is restricted in at least one dimension. Trapping is accomplished by providing a particular form of alternating electric potential which causes the particle to become trapped within a narrow 2D or small 3D volume. This volume can be made narrow (2D) or small (3D) enough to confine and stabilize microscopic and nanoscopic charged particles. The term “nanoscopic” and its variations indicate that the particle is of nanoscale dimensions (nanosized), i.e., a dimension sufficiently small that the properties of an object of such dimensions are predominantly governed by the behavior of individual atoms. Typically, a nanoscopic or nanoscale object refers to an object having at least one dimension within a range of about 1 to 100 nanometers (nm). The term “microscopic” and its variations indicate that the particle is of the dimensions of a micron. In some embodiments, the charged particle is suspended in a liquid environment. The liquid can be water or some other liquid such as glycerine. The liquid can also be a solvent. Electrolytes or other solutes may be present in the liquid. The charged particle to be trapped can be any particle of appropriate size and mass that is charged, including ions, molecules, polymers, nano- and micro-sized clusters. In some embodiments, the charged particle is a biomolecule. A biomolecule is any molecule that is involved in a biological process or found in a living organism. The biomolecule can be, for example, a nucleobase-containing molecule. Some examples of nucleobases include the pyrimidines (e.g., cytosine, thymine, and uracil) and the purines (e.g., adenine and guanine). Some examples of nucleobase-containing molecules that can be trapped herein include oligonucleotides, and nucleic acid polymers. The oligonucleotides and nucleic acid polymers can be deoxyribonucleic acid (DNA)-based or ribonucleic acid (RNA)-based. The biomolecule can also be, for example, an amino acid-containing molecule. Some examples of amino acid-containing molecules include the amino acids, peptides, oligopeptides, and polypeptides (e.g., proteins, such as enzymes). In other embodiments, the charged particle may be an inorganic molecule such as Silicon Dioxide. Silicon Nitride, or any type of nano or micro particle which is charged or may become charged in solution. Finally, nano and micron-size clusters, often show charging properties at the interface of their surface exposed to a liquid and electrolyte. A device which provides the alternating electric potential necessary for trapping a charged particle is called a Paul trap, which is a quadrupole type trap. Quadrupole trap types are those that lead to an electric potential F(x, y, z, t) of approximately quadrupolar spatial shape in the center. Their functionality emerges from the assumption that the particles are bound to an axis of the system if a binding force which acts on them increases linearly with their distance (F=−cr). Cylindrically symmetric electrical potential is ideally in the form of Φ ( r , z , t ) = Φ 0 2 r 0 2 ( α x 2 + β y 2 + γ z 2 ) . The condition that this potential has to fulfill the Laplace equation ∇Φ0=0 at every instant in time leads to a constraint α+β+γ=0 of the three geometric factors, which can be achieved in various ways, thus defining various possible geometries and types of quadrupole traps. From this constraint it follows that local three-dimensional minimum the potential can only trap charges in a dynamical way. The driving frequency and voltages can be chosen in such a way that the time-dependent potential will give rise to a stable, approximately harmonic motion of the trapped particle, in all or chosen directions. This can easily be demonstrated by a mechanical analogue. The equipotential lines form a saddle surface in a trap. A small, still ball set on the saddle is not in a stable equilibrium and will roll down the saddle. But if one sets the saddle into rotation with an appropriate frequency the ball motion will become stable in form of small oscillations and can remain on the saddle for an extended time. One of the most well known trap configurations is the 3D Paul trap, with α=β=1, γ=−2. This trap comprises ring-shaped metal electrode 1 and two cap-shaped metal electrodes 2, whose internal surfaces are defined as hyperbolic surfaces shown schematically in FIG. 1. The surfaces coincide with equipotential surfaces. Ring-shaped metal electrode 1 is halfway between the two cap-shaped metal electrodes 2, i.e. r02=2z02. The two cap-shaped metal electrodes 2 define a longitudinal axis between them along the z-direction. The ring-shaped metal electrode encircles this longitudinal axis in a plane transverse to the longitudinal axis. During operation, the charged particle is trapped in the space between these three electrodes by AC (rf oscillating, non-static) and DC (non oscillating, static) electric fields. A power source electrically connected between the two cap-shaped metal electrodes 2 and the ring-shaped electrode 1 provides the necessary voltage bias between the electrodes, as shown in FIG. 1. A charged particle located between the electrodes will experience a force due to the quadrupole electric potential. This force will cause the charged particle to become confined to a volume in the center of the trap significantly smaller than the dimensions of the trap itself. In the context of this invention, the term “trapping volume” and its variations refers to the volume to which a charged particle is localized after a short time. Such devices in a macroscopic scale have been widely fabricated and have proven to be a powerful tool in storage and detection of a single ion. Their typical dimensions are 100 μm to 1 cm, with voltages Vac in the range of 100-300 V, Vdc in the range of 0-50 V and the AC frequencies f=Ω/2π in the range of 100 kHz-100 MHz. If an electric bias of Φ0=Vdc−Vac cos(Ωt) is applied to the system in FIG. 1, the resulting azimuthally symmetric electric field is given by its components. E z = V dc - V ac cos ( Ω t ) z 0 2 z , E r = V dc - V ac cos ( Ω t ) 2 z 0 2 r Due to a periodic change in the sign of the electric force, one gets focusing and defocusing in both the r and z directions alternating in time. The equations of motion of a particle with mass M and charge Q in this field are given by Mathieu differential equations of motion, even in the presence of a damping force. This damping force, for example, can arise from collisional cooling in a gaseous environment or viscous forces in a liquid environment. If such a viscosity force is modeled by F=−Dν, where ν is the instantaneous velocity of the particle in the trap, and D is a constant proportional to the viscosity constant and geometrical features of the particle, the Mathieu equations of the damped motion of the particle, u=w exp(−kτ), are ⅆ 2 w ⅆ τ 2 + ( a - k 2 - 2 q cos ( 2 τ ) ) w = 0Where u stands for either the r or z coordinate, k = D / M Ω , τ = Ω t / 2 ,and a = 4 Q M V dc z 0 2 1 Ω 2 , q = 2 Q M V ac z 0 2 1 Ω 2 Here az=−2ar,qz=−2qz. The stability of the solutions to the equations, which defines the confining functions of the trap, is dependent on the values of parameters a, q, and k i.e. the stability depends on the magnitudes of both AC and DC components of the applied bias, on the angular frequency Ω, on the trap dimensions, on the particle charge Q and its mass M, as well as of the viscosity of the liquid, defined here by D. As a direct result of these stability parameters, it is possible to use a Paul trap to selectively trap particles. This is possible because different types of charged particles will vary in their charge to mass ratio Q/M. Because this characteristic influences stability, appropriately tuning parameters such as the Ω, Vdc, or Vac will create a trap that attracts one type of particle while repelling another type of particle. This property has is very desirable for sorting applications, including, for example, separating biomolecules in aqueous solutions. In the presence of a damping force, the regions in a-q-plane of the stable confinement are both enlarged and shifted in comparison to those with no damping. The solutions u may be bounded (stable) even if w is unbounded (unstable), due to the damping factor exp(−kτ). The effect of the collisions between the trapped particle and background particles will change the stable orbit of the particle, statistically increasing or decreasing its energy, which depends on the relative mass of the particles. For target particles larger than the background particles, such as biomolecules in water, this results in a damping force. The solution generally oscillates with a system “secular” frequency ω = β Ω 2 ,where β ≈ [ a - k 2 + q 2 / 2 1 - 3 q 2 / 8 ] 1 / 2 on which there is superimposed micromotion (of much higher frequencies Ω and 2Ω). Variations of the Paul trap involving more electrodes are possible. One common variation is the 2D Paul trap, which is also quadrupolar but with the parameters α=1=−β,γ=0. In this geometry, four confining electrodes are positioned around the trapping volume so that the quadrupole electric potential formed between them is orthogonal to the longitudinal axis of the trapping volume. One version of the 2D Paul trap, shown in FIG. 9A, is composed of cylindrically shaped electrodes extended in the z direction. This configuration provides trapping in the x and y directions without exerting a force in the z direction. Stable trapping in the x and y direction is determined by the values of a and q as determined above. FIG. 9B shows the regions in a-q-plane exhibiting stable and unstable confinement for such a 2D Paul trap in vacuum. In presence of liquid with various viscosity factors, the region of stability significantly changes and extends, as shown in FIGS. 9C and 9D. Trapping only in two dimensions can be beneficial because in some cases, such as trapping DNA, a trapping force exerted along the z axis may cause deleterious folding or rotation. Simulations of long DNA segments trapped in a 2D Paul trap have been performed in vacuum, the details of which can be found in sections three and four of Sony Joseph et al 2010 Nanotechnology 21 015103, incorporated herein by reference for all purposes. A polymer such as DNA behaves like a line charge within the trap, shown in FIG. 9A, and can be effectively trapped under conditions similar to a single charged particle of the same Q/M ratio. The DNA undergoes both oscillations and rotations in the trap depending on its initial angle, position and velocity, as well as of the angular bonding force of the adjacent atoms. These oscillations can be seen in FIGS. 15A-C. Free motion in the z direction in a 2D trap can be advantageously used for movement of the molecule into and out of the trap while maintaining a lateral confinement sufficient for measurement of particle properties. Motion of the charged particle in the z direction may be controlled by an appropriate DC field applied in the z direction. Alternating this DC field can allow the charged particle to be moved back and forth along the z axis. If a sensor is placed so that the charged particle passes by the sensor, such repeated trips past the sensor allow for multiple measurements and improvements in accuracy. Still further variations of the Paul trap include configurations which generate multipole potentials of a higher order than a quadrupole potential in the trapping volume. Multipoles of different orders include, for example, quadrupoles and octopoles. These potentials can be created by positioning more electrodes around the trapping volume. While the equations of motion are no longer given by Mathieu differential equations, the resulting alternating potential exhibits similar trapping capabilities. These multipole Paul traps can be thought of as a generalization of the basic Paul trap quadrupole configuration. In order to trap a charged particle, the trap must be at least a quadrupole. In some embodiments of the present invention the Paul trap is microscopic or nanoscopic. Reduced dimensions of the Paul trap allow for integration of the trap with other system components and provide trapping of charged particles in trapping volumes sufficiently small for detection or measurement of the charged particles. The trapping volume is a fraction of the size of the trap dimensions, typically a few percent of the size. This means that fabrication requirements are relaxed significantly compared to other trapping schemes in which the trapping volume is roughly the same size as the trap dimensions. The Paul trap can operate in liquid environments. In some embodiments, the liquid is a solvent containing an electrolyte. In a preferred embodiment, the liquid is an aqueous solution. This environment is particularly useful for trapping biomolecules, as many of these biomolecules become charged in such an aqueous solution. In other embodiments, the liquid may be glycerin or another viscous liquid. In the case of a Paul trap operating in a liquid environment, the space between the electrodes contains a liquid which carries the charged particle being trapped. In the context of this invention, the term “liquid container” and its variations refer this central portion of the trap when the Paul trap is adapted to operate in a liquid environment. The liquid container supports the liquid in the center of the trap so that the charged particle in the liquid can be trapped. The liquid container may comprise structures of the Paul trap such as the confining electrodes themselves or oxide layers which isolate the electrodes. The trapping volume is inside of the liquid container, so that a charged particle present in the liquid container can be pulled into and confined within the trapping volume. The liquid container may be fluidly connected to a microfluidic channel or other means for providing a liquid into the liquid container. During operation, the liquid is provided to the Paul trap. If the liquid container of the Paul trap is connected to a microfluidic channel, the liquid containing the charged particle to be trapped is provided through the microfluidic channel. Once the periodic bias is applied to the electrodes of the Paul trap, an electric potential is generated between the electrodes. Eventually, the charged particle will be enter the space between the electrodes and experience the electric potential. Due to the multipole form of this potential, the charged particle will move into the trapping volume and be trapped. Molecular Dynamics simulations of two nanoscale 3D Paul traps are shown in FIGS. 2-8 and illustrate the trapping function of the Paul trap on a Chlorine ion in vacuum and in an aqueous environment. These traps were of the 3D Paul trap configuration and constructed from gold atoms, with 2 nm diameter holes in the centers of each cap-shaped electrode to approximate an entrance and exit for the trapped ion. The two traps had dimensions, 2r0=5 nm, 2z0=5/√{square root over (2)} nm (nanotrap A), and 2r0=50 nm 2z0=50/√{square root over (2)} (nanotrap B). The parameters a and q were chosen in the middle of the stable region as defined for a conventional Paul trap. i.e. a=0.25 and q=0.4. Further details on the simulations can be found in sections two and three of Xiongce Zhao and Predrag S Krstic 2008 Nanotechnology 19 195702, incorporated herein by reference for all purposes. FIG. 2 shows coordinates of the ion as function of time in trap A at temperature of 3 K (corresponding to the ion energy of 2.5×10−4 eV). The initial coordinates shown have values (−12, 15, 24) Å relative to the geometric center of the trap, which were randomly set at the beginning of the simulation. The initial momentum of the ion was randomized following a Gaussian distribution but conformed to the system temperature. The needed trapping field is approximately Vdc=200 mV and Vac=600 mV, with the chosen frequency of AC voltage being 318 GHz. The trajectory of the ion was monitored for up to 3 ns of simulation. As can be seen in FIG. 2, the chlorine ion is driven to the center of the trap and rotated in a circular motion with its stabilized distance to the trap center being about 1.5 A. The time of 1.2 ns is elapsed before the stabilization is reached. The oscillation frequency of the trajectory is about 50 GHz, which is quantitatively consistent with the estimated “secular” frequency ω, for the given values of (a, q, k=0,Ω). Repeated runs lead to the same quantitative conclusions as the one shown, contributing to a statistical weight of the results. A series of simulations was performed by varying the driving fields from (Vdc=0.5 mV, Vac=1.5 mV) to (Vdc=200 mV, Vac=600 mV), and with frequencies ranging from 16-318 GHz. The AC voltage with a frequency in tens to hundreds of GHz is required in order to trap the charged ions within a timescale of nanoseconds. An increase in frequency, which also implies an increase in the magnitude of the voltage for given a and q, results in a faster establishment of stabilization. FIG. 3 shows simulations with AC voltage frequencies in the range of 159-318 GHz, which require voltages of Vdc=50 mV, Vac=150 mV to Vdc=200 mV, Vac=600 mV, at a constant temperature of 50 K (i.e. equivalent ion energy of 4.3×10−3 eV). The stabilization time for these systems ranges from 4.5 ns to 1 ns. However, the amplitude of the stabilized ion “secular” oscillations ranges from 12 Å to 6 Å, well below the dimensions of the trap. The stabilization time is dependent on the temperature, i.e. on the initial ion kinetic energy. FIG. 4 shows such variation of the initial ion energy in the range of 3-300K in trap A for Vdc=200 mV, Vac=600 mV, and f=318 GHz. There is an optimal temperature which yields the shortest stabilization time. For a chlorine ion under the above conditions the shortest stabilizing time occurs at 50 K, differing by almost a factor of 2 to the values at 3 K and 300 K. The oscillation amplitude of the ion inside the trap is also strongly dependent on the temperature. For example, at 300 K the ion is orbiting with a radius of 15 Å, whereas at 3 K the orbiting radius is only about 1.5 Å. The dependence of the orbiting radius is related to the temperature by the equation r2∝T. The results are plotted in FIG. 5. The simulated data point at 300 K does not overlap with the curve, simply because of the confinement effect of trap A. That is, the ion was not able to move beyond the trap cap along the z axis. Therefore, the effective circulating orbit of the ion at 300 K is depressed. In the 300 K case the orbiting trajectory of the trapped ion is changed slightly to adapt to the inner shape of the trap, although the circular nature of the orbit is not changed. In other words, the motion of the ion depends on the trap size. This phenomena will not be present for a macroscopic Paul trap but becomes significant when the trap is nanoscopic. This effect implies that a larger trap will tolerate an input ion with a higher energy, therefore higher temperature, without disturbing its orbiting motion in the z direction. FIG. 6 shows the trapping of a chlorine in trap B in vacuum. The increased dimensions of the trap allowed for lower trapping field frequencies, here chosen to be 20 GHz, and larger electric biases before a possible breakdown occurs. The ion was initially positioned at (−110, −100, 88) Å and the initial kinetic energy of the ion conforms to a system temperature of 300 K. The trapping fields were Vdc=80 mV, Vac=240 mV, which were turned on at t=0. As shown in FIG. 6, the ion was trapped in the center of the trap after a short time, with the orbital radius of about 6.5 nm. The overall behavior of the ion motion is similar to that observed in trap A, only with much bigger orbiting amplitudes. FIG. 7 shows a the trapping of a chlorine ion in trap A with an additional driving DC field along the z-axis (see FIG. 1) ranging from 10-150 mV/nm, which illustrates the impact of this additional field on the trapping process. Initially the ion was placed at the entrance of one of the cap holes with its coordinate as (5, −7, 22) Å relative to the trap center, with initial momentum of the ion set to conform to the system temperature as before. The z-direction driving field and the trapping fields were turned on simultaneously when the simulation started, and the trajectory of the ion was monitored for 3-12 ns. The near zero temperature of 3 K gives a clear picture of how the ion moves along each direction under the influence of the additional z-direction field. As seen in FIG. 7, the ion migrates through the central region from one entrance, while orbiting around the trap center in the x-y plane. The ion was finally stabilized at a position of about (0, 0, −21) Å, with the orbiting radius of about 2.1 nm. The ion is trapped similarly to a trap without the additional z-direction DC field although its position is now significantly shifted along the z-axis. The shift of the ion orbit along the z-axis varies with the strength of the field. Additional simulations indicate that the ion is stably trapped when the z field is below 110 mV/nm, while a field of 125 mV/nm would drive the ion all the way through the trap within 1 ns, without reaching stabilization. This suggests that for trap A the threshold driving DC field for moving the ion through the whole trap along z-direction lies between 110 to 125 mV/nm. By changing the polarity of the driving DC field it is possible to drive the ion back and forth along the z-axis through the trap. This back and forth movement provides an opportunity to increase the certainty of a measurement taken of the trapped ion inside the trap. FIG. 8 shows the trapping of a chlorine ion inside trap A filled with water, at 300 K with Vdc=4V, Vac=12V, f=80 GHz. Solvent polarization effects as well as impact from the collisions and thermal fluctuations of water molecules were treated through explicit atomistic MD simulations, using 5108 water molecules filling the volume of the trap. The trajectory of the ion shows that the stabilization process takes a much longer time than in vacuum even with stronger trapping electric fields. It takes about 12 ns for the ion to be trapped stably in the center of trap A. On the other hand, the ion experiences less fluctuations in movement during the stabilizing process along any of the directions, with a much smaller final oscillation amplitude of the ion in comparison to the same ion trapped in vacuum at the same temperature. One possible reason for such an effect is that the motion of the ion was effectively thermalized by water molecules around it due to the strong collision force from the electrostatic interactions. This shows the effects of background damping on the ion motion in the trap, suggesting that the addition of the solvent to the trap helps to stabilize the ion motion. FIG. 10A-13 show a 3D Paul trap implemented using conventional metal/insulator microfabrication approaches. This tri-layer crossing metal/insulator structure is used to form the Paul trap structure by etching nanopore 104 at the crossing of three metal electrodes. Nanopore has a diameter of approximately 20-50 nm, which is smaller than the widths of the three metal electrodes. Top electrode 101 and bottom electrode 102 act as the laterally confining electrodes. Middle electrode 103 forms the transversely confining ring electrode. A power source is electrically connected to the electrodes such that an AC and DC bias is applied between the top electrode and the middle electrode, and between the bottom electrode and the middle electrode. This AC and DC bias creates the time-varying periodic quadrupole electric potential necessary for trapping a charged particle. The walls of nanopore 104, including the electrodes and oxide insulator, form the liquid container. Trapping volume 106 is within the liquid container. The structure can be realized by a number of microfabrication approaches known in the art. The circular geometry necessary for a ring electrode necessary for the 3D Paul trap is achieved by a self-aligned etch through middle electrode 103. The top and bottom electrodes are realized by first creating the initial dielectric stack with the buried middle electrode, recessing from both sides a cylindrical etch pit (in which nanopore 104 will be centered), and backfilling with metal to make a pronounced quadrupole geometry. Middle electrode 103 is defined by a final etch through the structure creating the self-aligned translocation hole (nanopore 104) in the z-direction. Preferably, top electrode 101 and bottom electrode 102 have linewidths between 1-2 μm, with recesses 105 that are 100 nm deep. Middle electrode 103 is preferably 25-35 nm thick. Reactive Ion Etching techniques, known in the art, can be used to etch the nanopore through the stack. These preferred ranges demonstrate one of the key benefits of a Paul trap, which is that the trapping volume is much smaller than the device itself. This allows for nanoscale control using much larger device structures. In another embodiment implemented with microfabrication techniques, a 2D Paul trap is generated at the intersection of metal electrodes. As shown in FIG. 11A, a nanopore is etched at the intersection to form four coplanar electrodes 111. A power source electrically connected to the four coplanar electrodes 111 applies a voltage bias between adjacent electrodes. This voltage bias creates a quadrupole electric potential inside nanopore 112, providing confinement. With reference to FIG. 11B, the 2D configuration does not provide confinement along the axis of nanopore 112, which is orthogonal to the plane of the electrodes. Top electrode 113 and bottom electrode 114 provide control of the charged particle in this dimension. These electrodes may be initially pre-etched, added later, or electroplated to allow good definition of coplanar electrodes 111. A number of process variants can be used for coplanar electrodes 111, including straightforward nitride/metal definition/cap dielectric, followed by reactive ion etching of the through hole and metallization for top/bottom; or a Si process using SOI (Silicon-On-Insulator) and potentially implanted lateral electrodes. In a preferred embodiment, shown in FIG. 12A-B, a 2D Paul trap is constructed by patterning electrodes on a SiO2/Si wafer. Preferably, four Au/Cr (350/30 nm) electrodes are formed on top of this insulating substrate by UV-lithography and a double layer liftoff process. The tip to tip distance (2r0) for each pair of opposite electrodes ranges from 2-8 μm. Rather than making use of a nanopore, the liquid container in this preferred embodiment is simply the electrodes themselves and the substrate. An aqueous solution flows through a microfluidic channel. FIG. 12C shows the integration of the 2D Paul trap with the microfluidic channel. The microfluidic channel is formed by poly(dimethylsiloxane) (PDMS) using SU-8 as a molding master. Oxygen plasma treatment permanently bonds the PDMS onto the device surface and forms an anti-evaporation microfluidic channel. An inlet and an outlet are punched through before assembling. The device is wire-bonded and mounted onto a printed circuit board (PCB). Voltage in the form of U−V cos(Ωt), produced by a function generator (Tektronix AFG3252) together with a voltage amplifier (Tabor Electronics, Model 9250), is delivered to the device through 50Ω BNC cable and monitored by an oscilloscope (Tektronix DPO 4104). Since only the voltage difference between electrode pairs is of importance, for simplicity, only one phase AC potential is applied to one pair of electrodes while the other pair is referenced to zero and set to ground. However, a two channel mode with two out-of-phase signal (U−V cos(Ωt) and −U+V cos(Ωt)) is also used in some cases (mainly to get higher voltage difference). Particles in the microfluidic channel move freely in the x-y plane and into the liquid container within the trap, but are confined in the z direction by the channel height, which is controlled when fabricating the SU-8 master. The channel height is preferably less than 20 μm. FIG. 12D demonstrates the trapping functionality of the 2D Paul trap. The charged particle shown in FIG. 12D is a polystyrene bead in deionized water. The beads used had a mean radius of 490 nm and were functionalized with carboxylate groups (˜COOH) which give rise to the charge of the particle. An individual bead can be held in the trap for periods up to several hours. While trapping mostly occurs for single beads due to inter-particle Coulomb repulsion, multiple beads can be simultaneously trapped. In order to make use of this new technique for trapping charged particles in liquid environments, suitable methods for measuring properties of the trapped charges must be employed. In some embodiments, these measurements are optical measurements. In other embodiments, the measurements measure electronic properties of the charged particle. It is possible to directly observe particles confined within the Paul traps using an optical microscope. Observation can be aided by using fluorescently-tagged molecules or colloidal quantum dots. For example, a fluorophore-tagged DNA strand can be observed as it passes through the Paul trap using a fluorescent microscope. Optical measurements with greater spatial detail can be obtained using total internal reflection fluorescence microscopy (TIRFM). This technique employs the unique properties of an induced evanescent wave to selectively illuminate and excite fluorophores in a restricted specimen region immediately adjacent to a glass-water (or glass-buffer) interface. TIRFM requires an excitation light beam traveling at a high incident angle through a solid glass coverslip. Refractive index differences between the glass and water phases regulate how light is refracted or reflected at the interface as a function of incident angle. At a specific critical angle, the beam of light is totally reflected from the glass/water interface, rather than passing through. The reflection generates a very thin electromagnetic field (usually less than 200 nanometers) in the (aqueous) medium, which has an identical frequency to that of the incident light. This field, called the evanescent wave or field, undergoes exponential intensity decay with increasing distance from the surface. A schematic of the TIRFM principles is shown in FIG. 13. The characteristic distance for decay of the evanescent wave intensity is a function of the incident illumination angle, wavelength, and refractive index differences between media on each side of the interface. Fluorophores residing near the glass-liquid surface can be excited by the evanescent field, provided they have electronic transitions that occur in or very near the wavelength bandwidth. Fluorophores farther away from the surface avoid being excited, which leads to a dramatic reduction of unwanted secondary fluorescence emission from molecules that are not in the primary focal plane. The effect enables production of high-contrast images of surface events with a significant increase in signal-to-background ratio over classical widefield techniques. The penetration depth d, which usually ranges between 30-300 nm, decreases as the reflection angle grows larger. This value is also dependent upon the refractive indices of the media present at the interface and the illumination wavelength. In general, the maximum value of d is on the order of the incident wavelength. This provides a wide range of possibilities to control the evanescent field and a field intensity dependent tag response for the z-localization of the tagged molecule. In one embodiment, TIRFM is used in conjunction with a microfabricated Paul trap to observe fluorophore-tagged ssDNA. The microfabricated Paul trap is flip-chip mounted on a glass slide with surface-etched microfluidic channels for fluid flow, as shown in FIG. 14A. Fluid containing the fluorophore-tagged ssDNA flows into the liquid container 141 of the Paul trap. In this configuration the Paul trap is a 2D type where the liquid container is a nanopore. As the z-position of the tag is moved into and out of the evanescent field, the fluorescence is detected by a microscope on the optical axis. Because the excitation intensity is exponential with distance, the intensity versus the trap position setting can be measured. Nanoscale resolution is achieved, due to the lateral localization from the 2D Paul trap in conjunction with precise lateral position data from the exponential field intensity. Since evanescent fields propagate hundreds of nanometers in this system, it is possible to follow the ssDNA molecule all the way through the trap. An additional feature the trap system, not usual for TIRF microscopy, is that the longitudinal position of the particle can be independently tuned allowing for very high spatial resolution. A viable system for conducting TIRF measurements on trapped particles requires the evanescent field extend into the center of the trap. Because the range of the evanescent field is only 30-300 nm away from the glass, this presents difficulties with regards to mechanical stability, fluid flow resistance, and electrode lead fanout. In a preferred embodiment, shown in FIG. 14B, a pre-fabricated tiered glass mounting substrate is used. This structure allows the central trap region to be within the TIRF evanescent field while also allowing many microns of thickness for stability, fluid channels, and electrodes. Another class of measurements which can be done on trapped particles are electrical measurements. In nanoscale dimensions electronic properties such as capacitance and conductance are strongly influenced by the quantum mechanical structure of the molecules being measured. Because the atomic structures of molecules are different, under the right conditions they can be distinguished from one another. All references cited herein are incorporated herein by reference in their entirety and for all purposes to the same extent as if each individual publication or patent or patent application was specifically and individually indicated to be incorporated by reference in its entirety for all purposes. Many modifications and variations of this invention can be made without departing from its spirit and scope, as will be apparent to those skilled in the art. The specific embodiments described herein are offered by way of example only, and the invention is to be limited only by the terms of the appended claims, along with the full scope of equivalents to which such claims are entitled. |
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abstract | A system that measures the temperature distribution of the reactor coolant flowing through the hot leg or cold leg pipes by measuring the speed of sound time delay. This concept uses radiation hardened and temperature tolerant ultrasonic signal drivers based on vacuum micro-electronic technology. The system employs ultrasonic signals propagated through water, and relies on the characteristic that the speed of sound changes as the density and temperature of the water changes. Thus, a measured difference in the speed of sound in water may be directly correlated to a temperature change of the water. |
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054901847 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention is directed to a method and apparatus for determining on-line core wide power output of a pressurized water reactor (PWR) using the excore detector system, and to correction of the same for changes in three-dimensional power distribution in the core and in coolant density. 2. Background of Information The official determination of the reactor thermal power level in a PWR is based on a heat balance across the steam generators (S/G), called a secondary calorimetric measurement. The results of the calorimetric calculation are used to verify that the reactor is operating within the licensed reactor power level limits, and to calibrate the other indications of reactor power level. This calorimetric calculation is performed off line. Other indications of reactor power level, such as excore detector signal levels and RCS loop temperature indication values, are periodically calibrated against the calorimetric and used to provide on-line reactor power level inputs to the reactor control and protection systems. Unfortunately, errors in the reactor thermal power level calculation cause errors in all the other indications of reactor power. Presently, there is no easy way to detect and correct small errors in the reactor thermal power calculation. The major component of the heat balance calculation is the flowrate of feedwater into the S/G. The magnitude of the flowrate is typically determined using flow venturis installed in the main feedwater line for each S/G. The feedwater venturi flow readings are subject to systematic and random error mechanisms, which cause erroneous reactor power level calculation results. The most frequent systemic error, feedwater venturi fouling, causes the calculated reactor power to increase relative to the true reactor power level, necessitating a net reduction in the actual reactor power to maintain the apparent power within the operating limits. The amount of electrical output generated by the plant then decreases, reducing the revenue of the utility. There is a need, therefore, for an improved method and apparatus for on-line measurement of reactor power which is not affected by random errors in the thermal power measurements. There is also a need for a method and apparatus for correcting for errors in the calorimetric calculation of reactor power caused by feedwater venturi fouling or other types of systematic calorimetric error sources. SUMMARY OF THE INVENTION These needs and others have been satisfied by the invention which is directed to a method and apparatus for on-line determination of PWR power using excore detector signals corrected for changes in three-dimensional power distribution in the reactor core and for changes in coolant density. It has been found that feedwater venturi fouling is a phenomena that tends to increase with reactor operating time during a fuel cycle. The thermal power calculations performed early in the cycle tend to be relatively unaffected by venturi fouling effects, and therefore, tend to be quite accurate, allowing the power calibrations of the dependent power indications to be correspondingly accurate. In order to maintain the accuracy of the power calibration developed for the excore detectors in the absence of an accurate thermal power measurement, it is necessary to be able to normalize out the changes in the excore detector signals caused by changes in the core radial and axial power distribution that have occurred since the last accurate thermal power measurement was performed. It is also necessary to correct for the effects on indicated power caused by changes in coolant density that occur when reactor inlet temperature changes. This invention is directed to a method and system capable of allowing the signals from the excore detectors to be used as an independent means of determining the reactor power in an absolute fashion. In fact, the power measurements generated from the excore detectors in accordance with the invention can be used to ascertain the accuracy of and, if necessary, correct the thermal power measurements. In accordance with the present invention, excore detector current measurements are used to generate absolute reactor power by calibrating detector current measurements to the reactor thermal power calculation made at a base time early in the reactor cycle while the thermal reactor power measurement is still accurate. Measurements are also made at the base time of the three-dimensional core power distribution and the core inlet temperature distribution. Present core power measurements are then made by measuring the present excore detector current, the most recent three-dimensional core power distribution and the present core inlet temperature. The present core power is then calculated as the ratio of the present detector current to the detector current at the base time multiplied by the reactor thermal power measurement at the base time. The product is then corrected for changes in three-dimensional power distribution and in core inlet temperature since the measurement of those parameters at the base time. As the typical PWR excore detector system includes a plurality of excore detectors, usually four equally spaced around the reactor vessel, and each includes a top detector section and a bottom detector section, present core power determinations are made for each of the detector sections of each detector with the results averaged to determine present core power. The three-dimensional power distribution can be measured by an incore detector system which may utilize either fixed incore detectors or a moveable incore detector system. In the former case, a three dimensional core power distribution can be continually measured repetitively such as for instance every minute. The three-dimensional power distribution can also be generated by the system described in U.S. Pat. No. 4,774,049 which utilizes inlet temperature and readings from a pattern of thermocouples which measure core exit temperature to calculate on an on-line basis the three-dimensional core power distribution. The invention provides a simplified means for correcting the excore measurement for changes in coolant density. The correction factor is an exponential term in which the difference between the present core inlet temperature and core inlet temperature at the base time is multiplied by a constant. This constant is empirically determined at two different temperatures, preferably during reactor start-up. With the invention, only a single measurement of the thermal reactor power is required. A single calculation is made at a base time when the feedwater venturi used to measure the feedwater flow for the calorimetric calculation is unobstructed and the thermal reactor power calculation is accurate. The invention embraces both the method and apparatus for absolute excore detector reactor power determination. |
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042343844 | description | DETAILED DESCRIPTION OF THE INVENTION As disclosed, generally FIG. 1 and, in greater detail, in FIG. 3, the core 1 of a high temperature gas cooled nuclear reactor is formed by a bed of spherical fuel elements 2 and is generally surrounded by an annular side reflector 3. The reflector 3 is constructed of a plurality of graphite bricks 4. Removal tubes 5 are illustrated at the bottom of the bed of spherical fuel elements hereinafter referred to as the pebble bed. Six of these removal tubes lead from the pebble bed to a charging installation (not illustrated). In FIG. 2 the regular arrangement of the pebble bed removal tubes 5 is illustrated on a pit circle around the axis of the reactor core. For each pebble removal tube 5, a conical pebble inlet 6 is provided. The inlet itselt is formed by part of the support structure. The support structure consists of three layers of prismatic graphite blocks constructed as a closed unit without expansion gaps, as illustrated in FIG. 2. The top layer 7 and the intermediate layer 8 are composed of a plurality of hexagonal graphite blocks 9 connected through a keying arrangement 10 with each other. The graphite blocks 9 of the top layer 7 and intermediate layer 8 are designed with respect to their height so that the conical pebble inlets 6 are necessarily formed, as illustrated in FIG. 3. The bottom layer 11 of the support structure is formed by a number of support units 12. These units are also interkeyed with the intermediate layer 8 which is located above them. Beneath bottom layer 11, a hot gas collector space 13 is provided for collecting cooling gas. The cooling gas flowing through the fuel element pebble bed from top to bottom collects in this space. In the area of the pebble removal tubes 5, support units 12 are designed so that no gaps remain between the pebble removal tubes and the various support units. The support units 12 and the graphite blocks 9 again exhibit different cross sections toward the side reflector 3. Beneath the side reflector 3, the hot gas collector space 13 is in direct communication with an expanding area 14 making possible the connection of the radial cooling gas conduits with the collector space. As illustrated in FIG. 5, each support unit 12 consists of several support segments 15 fitted together into a hexagonal cross section. At the location of lower layer 11 representing the juncture of three support units 12, cooling gas channels 16 are provided. These terminate in the hot gas collector space 13. All of the support units 12 rest on a round column 17, directly through a column head 18. The column head 18 is set into a recess located in the center of the support unit and interkeyed with it. The round columns 17 traverse the hot gas collector space 13 and adjoin to the bottom of the nuclear reactor. As illustrated in FIG. 2, the support units 12 are of a hexagonal cross section and are significantly larger than the hexagonal graphite blocks 9 of the uppermost and intermediate layer of the support structure. Each support unit must, according to the invention, carrying of several graphite blocks 9 and of the fuel element pebbles 2 resting on the graphite blocks. In the top layer 7 and intermediate layer 8, a plurality of "central" graphite blocks 19 is present. The central blocks 19 are always surrounded by six "peripheral" graphite blocks 20 as illustrated in FIG. 4. Each of the peripheral graphite blocks 20 forms a part of the boundary of three different central graphite blocks 19. In this manner, the group comprising seven graphite blocks are each boxed in with each other over the layer 11. Each of the central graphite blocks 19 is aligned with one of the round columns 17. In all of the hexagonal graphite blocks 9 of the uppermost layer 7, a plurality of small vertical borings 21 are provided for the cooling gas. These are connected with the collector spaces 22 present in the graphite blocks 9 of the intermediate layer 8, while the collector space is located in the central graphite block 19 are designed as sack-like borings 22A. The collector spaces in the peripheral blocks 2 present continuous borings 22B. These borings are aligned with a cooling gas channel 16 in the support units 12. The sack-like borings 22A in the central graphite blocks 19 are connected with the continuous borings 22B connected with several connecting borings 23 as illustrated in FIGS. 3 and 4. In FIG. 3 one of the connecting borings 23 is rotated in the illustration. Furthermore, FIG. 3 illustrates the way by which differential expansion of the side reflector 3 in the three layers, 7, 8 and 11, of the support structure is permitted in the continuous vertical separating gap 24. The drawings are illustrative of the preferred embodiment of the present invention and are not intended to limit the disclosure of the present invention in any unduly restrictive manner. Obvious equivalents will be recognized as suitable to the skilled artisan and are considered to be included in the concept of the invention. |
claims | 1. An X-ray converter comprising:a light-proof housing, one of whose walls is X-ray transparent, and the following units fastened one after another behind this wall:an X-ray-to-optical converter of the X-radiation into visible light,a filter of residual X-radiation,an unit of objective lenses, each of which contains at least two one by one installed lenses for focusing a part of the light flux on the corresponding optoelectronic converter, anda photodetector containing at least two optoelectronic converters having partly overlapping fields of view and separated electrical outputs for connection to a system for fragmentary video signals processing and their “sewing together” into an integral output video signal, characterized in thatthe light-proof housing is equipped with an additional light-opaque and X-ray-opaque partition that has through-holes, whose number and placement correspond to the number and placement of objective lenses and optoelectronic converters, and is rigidly fastened within said housing practically parallel to the X-ray-to-optical converter,the filter of residual X-radiation formed as washers that are made from an X-ray opaque light-transparent material and rigidly fastened within said through-holes of the additional partition,said additional partition is equipped with blinds, whose number and placement correspond to the number and placement of objective lenses and optoelectronic converters; these blinds are installed on such side of this partition that is opposite to said X-ray-to-optical converter, andlength A of each blind and distance D from the front (following the pass of X-rays) surface of said X-ray-to-optical converter to the plane of front (following the pass of light) end faces of objective lenses are related by the ratio A/D=(0.50-0.95). 2. The X-ray converter of claim 1 characterized in that said ratio is of A/D=(0.55-0.90). 3. The X-ray converter of claim 2 characterized in that said additional partition comprises of a lead plate and a supporting plate made from suitable rigid material. 4. The X-ray converter of claim 3 characterized in that each objective lens is equipped with at least one diaphragm for restriction of the light flux. 5. The X-ray converter of claim 4 characterized in that each objective lens has three diaphragms installed ahead of the input lens, between lenses, and after the output lens. 6. The X-ray converter of claim 2 characterized in that surfaces of said washers of the filter of residual X-radiation and the lenses of said objective lenses have anti-reflecting coatings. 7. The X-ray converter of claim 1 characterized in that said additional partition comprises of a lead plate and a supporting plate made from suitable rigid material. 8. The X-ray converter of claim 7 characterized in that each objective lens is equipped with at least one diaphragm for restriction of the light flux. 9. The X-ray converter of claim 8 characterized in that each objective lens has three diaphragms installed ahead of the input lens, between lenses, and after the output lens. 10. The X-ray converter of claim 9 characterized in that interior of the side housing walls, blinds on the inside, and diaphragms on both sides have black matte coatings. 11. The X-ray converter of claim 9 characterized in that surfaces of said washers of the filter of residual X-radiation and the lenses of said objective lenses have anti-reflecting coatings. 12. The X-ray converter of claim 8 characterized in that interior of the side housing walls, blinds on the inside, and diaphragms on both sides have black matte coatings. 13. The X-ray converter of claim 12 characterized in that surfaces of said washers of the filter of residual X-radiation and the lenses of said objective lenses have anti-reflecting coatings. 14. The X-ray converter of claim 8 characterized in that surfaces of said washers of the filter of residual X-radiation and the lenses of said objective lenses have anti-reflecting coatings. 15. The X-ray converter of claim 7 characterized in that surfaces of said washers of the filter of residual X-radiation and the lenses of said objective lenses have anti-reflecting coatings. 16. The X-ray converter of claim 1 characterized in that surfaces of said washers of the filter of residual X-radiation and the lenses of said objective lenses have anti-reflecting coatings. |
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041815718 | abstract | A fuel pin bracing grid for a nuclear fuel sub-assembly comprises a honeycomb array of unit cells formed from discrete strips. The cells are hexagonal, three alternate sides having windows and the remaining sides have linear groups of three embossments to provide guide pads for fuel pins. The openings provide a measure of compliancy for the grid to facilitate insertion and withdrawal of the pins. A fuel sub-assembly for a liquid metal cooled fast breeder nuclear reactor has a central fuel section with end extensions, the fuel section comprising a bundle of fuel pins in a hexagonal wrapper the pins being braced by a series of grids according to the invention. Reprocessing of the fuel is facilitated because the pins are withdrawable collectively from the compliant grids and wrapper combination merely by cutting an end extension from the wrapper. |
summary | ||
summary | ||
062326791 | description | BEST MODE FOR CARRYING OUT THE INVENTION Referring now to the Figure, an illustrative embodiment of the invention is shown in a heating and electricity generating system 10. The heating and electricity generating system 10 comprises a generator 20 for generating electricity, a pump 30 for pressurizing liquified working fluid, a pump 40 for circulating engine coolant, an oil fired heater 50, a turbine 60 for driving generator 20 through shaft 62 and a condenser 70 for liquefying the working fluid. A control 80 receives power from generator 20 through conductor 28 and supplies power to power consumers through conductor 88. Control 80 also controls pump 30, pump 40 and oil fired heater 50 through means indicated by dashed lines 82, 84 and 86 respectively. It will be appreciated as the description proceeds that the invention may be implemented in different embodiments. Referring now to FIG. 1, the heating and electricity generating system 10 comprises a generator 20 which can be any of many known designs. The preferred generator design is a high speed generator with a permanently magnetized armature and electronic commutation of the stator coils. This design is known to have the advantages of small size and high efficiency. The pump 30 may be of any known design but is preferably a piston or gerotor type pump powered by an electric motor. The pump 40 is preferably a centrifugal pump powered by an electric motor designed to provide a flow of several gallons per minute of engine coolant. The oil fired heater 50 may be of any conventional design. A preferred design uses an oil burner of the catalytic type commonly used in room heaters. Means such as fire tubes are provide to transfer heat from the combustion gases produced by the oil burner of heater 50 to the liquified working fluid. The turbine 60 is preferably a radial inlet centrifugal turbine but may be of any known design suitable for converting pressurized working fluid provided by oil fired heater 50 to mechanical energy. Computer codes are available from many sources for computing the exact shape of a suitable turbine rotor and housing depending on the working fluid, operating temperatures and power rating. The turbine rotor is preferably made by an inexpensive process such as casting and is preferably made of an inexpensive metal such as a steel. Condenser 70 receives spent working fluid from turbine 60 and cools it to liquify it and transfers the heat it receives to engine coolant provided by pump 40. Condenser 70 may be of any design known to be suitable by those skilled in the design of heat exchangers. Control 80 operates to control power to the coils in generator 20 if required which would be required if generator 20 is of the type having a rotor comprising permanent magnets. Control 80 also operates to convert the power it receives from generator 20 to a desired form such as 120 volt 60 Hertz ac or 12 volt dc and deliver that power through electrical conductors 88. Control 80 also receives from a switch 90 located in the vehicle cabin a signal turning the heating and electricity generating system 10 on or off. Control 80 also indicates to pumps 30 and 40 and oil fired heater 50 by means indicated by dashed lines 82, 84 and 86 respectively that they should turn on or off. Heating and electricity generating system 10 is a sealed system filled with a working fluid. The working fluid is selected to have the desirable properties of low sound velocity in the vapor, high heat of vaporization, a suitable condensation temperature at the operating pressures, a high critical temperature, a low freezing temperature, a low viscosity and a high film heat conductivity. An optimum working fluid has not been identified but decane and is believed to be suitable. A primary consideration is that the low temperature of the cycle is likely to be the temperature of a hot engine and the working fluid must have a low vapor pressure at these high temperatures so it can be condensed in condenser 70. The operation of the heating and electricity generating system 10 of this invention will now be described with reference to FIG. 1. In operation of the system, when the truck operator closes switch 90, control 80 initiates operation of the heating and electricity generating system 10 by providing power or otherwise controlling pump 30, pump 40, and oil fired heater 50 to begin operating. Pump 30 pumps any liquid working fluid in condenser 70 through fluid conduit 32 into oil fired heater 50. Pump 40 circulates engine coolant through condenser 70 for withdrawing heat from the working fluid and causing it to condense to a liquid. Oil fired heater 50 burns diesel fuel from the main fuel supply of the truck to heat and vaporize and thereby pressurize the working fluid. Preferably, the working fluid vapor is then superheated by oil fired heater 50. The superheated working fluid vapor is conducted by conduit 52 to the inlet of turbine 60 where it causes turbine 60 to turn shaft 62 and the armature of electric generator 20. The spent working fluid vapor (with any entrained liquid) is conducted by fluid conductor 62 to condenser 70 where the spent working fluid condenses to a liquid and the heat it contains is transferred to the engine coolant. Heat in the coolant is carried to the heater core in the passenger compartment in cold weather to heat the passenger compartment and in warm weather it is sent directly to the engine block which operates as a large heat sink. Generator 20 generates electrica power which is supplied through conductor 28, control 80 and conductor 88 for use by the electrical appliances in the truck cabin. An advantage the heating and electricity generating system 10 is that it is acoustically quiet because all of the components when combined as described hereinabove are not noisy. More particularly, the pump 30 is quiet by virtue of its small size. Pump 40 is also quiet by virtue of being a centrifugal pump of small size and low pressure. Oil fired heaters of the catalytic type are commercially available in designs that are acoustically quiet. Turbine 60 is also inherently quiet. Therefore the cogeneration system described hereinabove can be installed in such as the engine compartment of a truck or elsewhere outside the truck cabin and will not generate objectionable sound even when the truck is parked and the operator is sleeping. By comparison, other low cost cogeneration plants are undesirably noisy without sound insulation means. A particular example of a cogeneration plant that is undesirably noisy comprises a Diesel engine driving an electric generator. A second advantage of the heating and electricity generating system 10 is that it minimizes pollution. Catalytic heaters produce sufficiently low amounts of objectionable gases that they are commonly sold for use inside residences. A third advantage of the heating and electricity generating system 10 is that all components are likely to last for the lifetime of the truck without maintenance. Gerotor pumps are used as oil pumps in engines and last the lifetime of the engine. Centrifugal pumps can be designed to last far longer than required. Oil fired heaters are simple and reliable. Turbines on fluid bearings have no wear except at the bearings which can be designed to provide the desired lifetime. Although the description of this invention has been given with reference to a particular embodiment, it is not to be construed in a limiting sense. Many variations and modifications will now occur to those skilled in the art. For a definition of the invention reference is made to the appended claims. |
048759457 | abstract | The exhaust gas of a fusion reactor contains, besides non-burnt fuel (tritium and deuterium) and helium, the "ash" from the nuclear fusion reaction a number of impurities with the radioactive tritium and/or deuterium chemically bound to them. In order to clean the exhaust gas, both the elemental and the chemically bound tritium and/or deuterium fractions are separated from the exhaust gas. Separation is achieved exclusively by physical and catalytical process steps, namely a palladium/silver permeator, a CuO/Cr.sub.2 O.sub.3 /ZnO catalyst bed and a further palladium/silver permeator containing a nickel/aluminum oxide bulk catalyst. |
046719237 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The basic components of a holddown spring retention assembly include lift lugs 2, one of which is shown in FIG. 1, and hangers 4, one of which is shown in FIG. 2. Each lift lug 2 includes a support plate portion 6 and two projecting plate portions 8 located at opposite edges of portion 6 and projecting at right angles thereto. Support plate portion 2 is provided with studs, or posts 10 and bores 12 (shown in FIG. 3) for mounting lug 2 on a holddown spring 14, a portion of which is also shown in FIG. 3. Studs 10 may be welded to plate portion 6. Each hanger 4 has the general form of an I-beam having an upright web portion 16 and upper and lower flange portions 18 and 20. Hanger 4 further includes a generally trapezoidal portion integrally machined with web portion 16 and lower flange portion 20. Hanger 4 further has shoulder portions 26 enclosed by, and integral with, web portion 16, lower flange portion 20 and plate portion 24, one shoulder portion being located to either side of web portion 16. Upper flange portion 18 is provided with through bores 28 for securing hanger 4 in position, in a manner to be described below. A complete assembly according to the invention can be composed of three lift lugs 2 and hangers 4, each hanger being associated with a respective lug, and each lug being permanently fastened to the outer periphery of holddown spring 14 by means of threaded bolts extending through studs 10 and pins extending through bores 12, the mounting of one lug 2 on spring 14 being shown in FIG. 3. Spring 14 is a solid, annular, Bellevilletype spring having three lift lugs 2 secured to its outer periphery with a mutual spacing of 120.degree. about the circumference of spring 14. The present invention is primarily directed to a reactor which is so constructed that holddown spring 14 will be installed between the upper flange of the core barrel and the upper flange of the reactor upper internals inner barrel. In this case, and as shown in FIG. 4, each hanger 4 is mounted in a radially extending, generally T-shaped recess or groove 30 extending radially inwardly from the outer circumference of the upper flange 32 of the reactor upper intervals inner barrel. The upper flange portion 18 of hanger 4 is lodged in the cross-arm of the "T," while the web portion 16 extends along the vertical part of the "T." Hanger 4 is bolted in place by means of bolts 34, one of which is shown in FIG. 5, passing through bores 28 and screwed into threaded blind bores in upper flange 32. Lower flange portion 20, plate portion 24 and shoulder portions 26 of hanger 4 are then suspended below upper flange 32. After hangers 4 have been assembled with lift lugs 2, in a manner to be described below, hangers 4 have been bolted in place in grooves 30, and the inner barrel has been installed in the reactor vessel, an upper calandria is installed into the vessel so that the flange 36 of the upper calandria rests upon upper flange 32 and covers each upper flange portion 18 and its associated bolts 34. The complete assembly of a holddown spring and its associated lift lugs and hangers is shown in the cross-sectional detail view of FIG. 5. As can be seen, spring 14 rests upon the upper surface of the upper flange 38 of a core barrel 40 constituting a component of the reactor. The upper flange 32 of reactor upper internals inner barrel 42, in turn, rests upon spring 14, and, as mentioned above, the flange 36 of upper calandria 44 rests upon flange 32. Flange 38 of core barrel 40 is supported on a radially inwardly extending shoulder formed at the inner wall of the reactor core pressure vessel 46. FIG. 5 illustrates one lift lug-hanger unit according to the invention, it being understood that a complete spring retention assembly according to the invention will include a plurality of, typically three, such units equispaced about the circumference of flange 32. As shown, the lift lug is secured to the outer periphery of spring 14 so that, in the assembled condition of the reactor components, projecting plate portions 8 are disposed above, and spaced slightly from, lower flange portion 20 of the hanger. The hanger is suspended from flange 32 and is secured thereto by means of bolts 34. The cross section of spring 14 includes a lower annular boss adjacent the inner periphery of spring 14 and bearing against the upper surface of flange 38 and an upper annular boss adjacent the outer periphery of ring 14 and bearing against the lower surface of flange 32. In the installed state illustrated in FIG. 5, the presence of these bosses will cause spring 14 to pivot about a horizontal circular axis in order to prestress spring 14 and create the desired clamping action. In order to enable spring 14 to undergo this movement, projecting plate portions 8 of the lift lug are dimensioned to be spaced slightly above the associated lower flange portion 20 when the reactor components are in the assembled state shown in FIG. 5. The spring retention assembly according to the invention comes into operation when it is desired to remove inner barrel 42 together with holddown spring 14 for inspection, maintenance, or repair purposes. At that time, when inner barrel 42 is lifted, in the usual manner, the plate portions 8 of lift lugs 2 will come to rest upon lower flange portions 20, so that spring 14 will be automatically lifted with barrel 42, while retaining its proper radial orientation. Therefore, when barrel 42 is reinstalled, spring 14 will automatically be properly positioned. Initial installation of the spring retention assembly according to the invention can be effected in the following manner: Before the initial installation, lift lugs 2 are secured to spring 14 at the appropriate locations. As the first installation step, core barrel 40 is installed in pressure vessel 46 with flange 38 resting on the radially inwardly extending shoulder formed at the inner wall of vessel 46. Then a specially constructed standoff (not shown) is placed upon the upper surface of flange 38. This standoff may be composed of a plurality of, e.g., six, individual arcuate segments which will be spaced about the circumference of flange 38. The standoff is constructed to be in vertical alignment with the intended location of spring 14 and the standoff is dimensioned so that when it rests on flange 38 it presents an upper support surface which projects above the upper edge of vessel 46. Then eyebolts are secured in threaded bores provided in the upper surface of spring 14, one of which is shown in FIG. 5, and a lift sling is hooked onto those eyebolts. Spring 14 can then be lowered onto the standoff 38. The eyebolts can then be removed from spring 14. Then, inner barrel 42 is placed on spring 14 so that flange 32 rests upon spring 14. Then, each hanger 4 is installed by sliding upper flange portion 18 into the upper horizontal part of groove 30 so that lower flange portion 20 comes to lie below the associated projecting plate portions 8. Thereafter, bolts 34 are installed to permanently secure each hanger to flange 32. Once these steps have been performed, it will not be necessary to remove the hangers during the operating life of the reactor. Finally, the assembly of barrel 42 and spring 14 is lifted by an amount necessary to permit removal of the standoff segments. Then this assembly can be lowered into its final position with spring 14 resting upon the upper surface of flange 38. Subsequently, upper calandria 44 is installed in pressure vessel 46 and assembly can be completed in the usual manner. At various times during the operating lifetime of such a reactor, it is necessary to gain access to the region of the upper surface of flange 38 which is covered by spring 14. For example, in the proposed new reactor design, this region will be provided with access holes for radiation specimens installed in core barrel 40 and rotolocks which are used for lifting core barrel 40 out of pressure vessel 46. With the holddown assembly according to the invention, spring 14 will be automatically removed together with inner barrel 42 and will automatically assume its desired position upon reinstallation of barrel 42. If spring 14 could not be lifted together with barrel 42, it would be necessary to either remove that spring by reinstalling the eyebolts used during initial installation, which is an extremely difficult and time-consuming task when spring 14 is within pressure vessel 46, or it would probably be necessary to mount spring 14 between flanges 32 and 36, which presents several disadvantages such as requiring higher spring loads and lower reliability. Proper radial positioning of spring 14 during reinstallation is controlled by dimensioning lift lugs 2 and hangers 4 so that there is a sufficiently small radial clearance between the outer edges of projecting plate portions 8 and the interior surface of plate portion 24. Preferably, a clearance of less than 5 mm is provided, with respect to the dimensions of the various components at room temperature. Circumferential positioning of spring 14 relative to flange 22 is maintained by appropriate selection of the circumferential clearance between projecting plate portions 8 and shoulder portions 26. Preferably, a clearance of the order of 1.25 cm is provided between each shoulder portion 26 and plate portion 8 when lug 2 is centered on hanger 4. Such a clearance further assures that plate portions 8 will not shift circumferentially beyond the edges of lower flange portions 20. A second embodiment of the invention is shown in FIGS. 6 and 7. Each lift lug 52 is provided with a support plate portion 56 secured to spring 14 and carrying a single, centrally located, projecting plate portion 58. Lug 52 is secured to spring 14 by threaded bolts 60 extending through studs 10, two studs 10 being disposed to each side of plate portion 58, and by pins 62 passing through bores 10. Spring 14 is shown in FIG. 6 but not in FIG. 7. Hanger 62 has the form of a "C" when viewed in a radial direction, i.e. hanger 62 opens to one side in the direction of the circumference of spring 14. Hanger 62 is composed of a vertical web portion 66, an upper flange portion 68 secured at one end to the top of web portion 60 and a lower flange portion 70 secured at one end to the bottom of web portion 66. Lower flange portion 70 supports projecting plate portion 58 when spring 14 is being lifted with the reactor upper internals inner barrel upper flange 72 (shown in FIG. 7); when all components are installed in the reactor pressure vessel, as shown in FIGS. 6 and 7, projecting plate portion 58 is spaced a small distance above lower flange portion 70. In the installed state, the clamping force on spring 14 causes lug 52 to be tilted slight, as shown in FIG. 6. Upper flange 72 is provided with a generally horizontal groove 74 extending to both its outer periphery and upper surface for receiving upper flange portion 68 and upper flange portion 68 is bolted to upper flange 72 by a bolt (not shown) secured in countersunk bore 76. Upper flange 72 is further provided with a slot 78 communicating with groove 74, extending to the outer periphery of upper flange 72 and extending across the full height of upper flange 72. Slot 78 is dimensioned to permit passage of hanger 62 so that the hanger can be installed after spring 14 and the upper internals inner barrel have been installed in the reactor pressure vessel. Thus, a standoff is not required and installation of the holddown spring retention assembly is greatly simplified. Hanger 62 can even be replaced, if necessary, without removing the inner barrel from the reactor pressure vessel, i.e. only the upper calandia need be removed for this purpose. In order to assure proper circumferential positioning of spring 14 relative to flange 72, and to keep plate portions 58 from slipping off lower flange portions 70, one of the hangers 62 is oriented to face circumferentially in the opposite direction to that shown in FIG. 7, so that, for example, two hangers 62 open circumferentially in the clockwise direction about spring 14 while the third hanger 62 opens circumferentially in the counterclockwise direction. In the embodiment shown in FIGS. 8 and 9, left lug 82 has the same general form as lift lug 52, the only difference being that projecting plate portion 84 has a smaller radial dimension than does projecting plate portion 58. Therefore lift lug 82 will not be described further. Hanger 86 again has the form of a "C", but this time when viewed along the circumference of spring 14, and is composed of a vertical web portion 88, an upper flange portion 90 secured at one end to the top of web portion 88 and a lower flange portion 92 secured at one end to the bottom of web portion 88. Here again, lower flange portion 92 supports projecting plate portion 84 during lifting of spring 14 and is spaced therefrom in the assembled state of the reactor. Web portion 88 is spaced radially from projecting plate portion 84 by a distance selected to assure radial centering of spring 14. To provide proper circumferential positioning of spring 14, hanger 86 is provided with two abutment members 94 welded to vertical web portion 88 and spaced apart in the circumferential direction by a distance selected to provide the desired clearance to each side of projecting plate portion 84. Hanger 86 is secured to inner barrel upper flange 96 by two bolts inserted in countersunk bores 98 and flange 95 is provided with a groove extending inwardly from its periphery and having a horizontal groove portion for receiving upper flange portion 90 and a vertical groove portion for receiving the upper part of vertical web portion 88. Assembly of the embodiment of FIGS. 8 and 9 can be effected in the manner described earlier with reference to the embodiment of FIGS. 1-5. After assembly, the projecting portion 84 of each lug 82 will be located between abutment members 94 of the associated hanger 86 and will be positioned radially by vertical web portions 88. All of the disclosed embodiments offer the advantage that they will not interfere with normal thermal expansion of the reactor components since, with increasing temperature, the reactor internals inner barrel upper flange will always undergo a greater radial outward displacement than will spring 14. Therefore, the various surfaces of each hanger will move away from the associated surface of each lift lug. In addition, when the reactor is assembled, each lift lug is completely out of contact with its associated hanger. It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptations, and the same are intended to be comprehended within the meaning and range of equivalents of the appended claims. |
047568780 | abstract | A marked reduction in the pressure drop of cooling liquid through a grid spacer of a nuclear fuel assembly is attained by convexly contouring the upstream (usually the lower) edges of the grid members. Preferably, they are made streamlined or semicylindrical. This can be done by first beveling and then etching them, by directing a stream of mixture of abrasive and an organic polymer against them, or by traversing an electron or laser beam along them at a power and velocity such as to cause local melting. A lesser improvement is secured by beveling alone. A still further improvement can be obtained by also tapering the downstream (usually upper) edges. |
claims | 1. A method, comprising:placing into a crucible precursors including:a rare earth halide precursor; anda Ca halide precursor;melting the precursors to form a melt, wherein a concentration of the Ca halide is at least 0. 02wt. %;forming a scintillation crystal including Ln(1-y)REyX3:Ca, wherein:Ln represents a rare earth element;RE represents a different rare earth element;y has a value in a range of 0 to 1; andX represents a halogen; andoptically coupling an optical interface to the scintillation crystal. 2. The method of claim 1, wherein RE is Ce. 3. The method of claim 2, wherein Ln is La. 4. The method of claim 3, wherein X is Br. 5. The method of claim 2, wherein y is 1.0 f.u. 6. The method of claim 1, wherein the concentration of the Ca halide in the melt is no greater than 1.0 wt. %. 7. The method of claim 1, wherein y is no greater than 0.5 and at least 0. 005. 8. The method of claim 1, wherein y is in a range of 0.01 to 0.09. 9. The method of claim 1, wherein for a radiation energy range of 13 keV to 30 keV, the scintillation crystal has a PR dev average of no greater than 14%, or for a radiation energy range of 30 keV to 60 keV, the scintillation crystal has a PR dev average of no greater than 8.0%. 10. The method of claim 1, wherein:for a radiation energy range of 11 keV to 30 keV, the scintillation crystal has a PR dev average of no greater than 8.0%; orfor radiation energy range of 30 keV to 60 keV, the scintillation crystal has the PR dev average of no greater than 3.6%. 11. The method of claim 1, wherein an energy resolution ratio is an energy resolution of the scintillation crystal divided by a different energy resolution of a different scintillation crystal of a different composition, wherein the energy resolution ratio is:no greater than 0.95 for an energy of 8 keV;no greater than 0.95 for an energy of 13 keV;no greater than 0.95 for an energy of 17 keV;no greater than 0.95 for an energy of 22 keV;no greater than 0.95 for an energy of 26 keV;no greater than 0.95 for an energy of 32 keV; orno greater than 0.97 for an energy of 44 keV. 12. The method of claim 11, wherein the energy resolution ratio is no greater than 0.95 for the energy of 8 keV. 13. The method of claim 11, wherein the energy resolution ratio is no greater than 0.95 for the energy of 13 keV. 14. The method of claim 11, wherein the energy resolution ratio is no greater than 0.95 for the energy of 17 keV. 15. The method of claim 11, wherein the energy resolution ratio is no greater than 0.95 for the energy of 22 keV. 16. The method of claim 11, wherein the energy resolution ratio is no greater than 0.95 for the energy of 26 keV. 17. The method of claim 11, wherein an energy resolution ratio is no greater than 0.95 for the energy of 32 keV. 18. The method of claim 1, further comprising optically coupling a photosensor to the optical interface. |
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abstract | A blade device for forming a hollow cone-like radiation includes a pair of blade members opposed to each other symmetrically at a fixed inclination angle in a first direction perpendicular to the axis of the cone, a pair of second blade members opposed to each other symmetrically at a variable inclination angle in a second direction perpendicular to the axis of the cone and also perpendicular to the first direction, a pair of lever members fixed respectively at one ends thereof to faces of the pair of second blade members on the side opposite to the mutually confronting side, and lever actuating unit for pivoting the pair of lever members about respective shafts. |
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summary | ||
description | 1. Technical Field The present invention relates generally to ion implantation, and more particularly, to a system, method and program product for determining parallelism of an ion beam using a refraction method. 2. Related Art Controlling the angle or parallelism of an ion beam is important for the proper operation of various different types of devices and processes. Ion implantation is a standard technique for introducing conductivity-altering impurities into, or doping, semiconductor wafers. A typical ion implantation process uses an energetic ion beam to introduce impurities into work pieces, i.e., semiconductor wafers. As is well known, introducing the impurities at a uniform depth and dose into the work pieces is important to ensure that semiconductor devices being formed to operate properly. FIG. 1 shows schematically, in three dimensions, a conventional implantation of an ion beam into a wafer. Z-Axis and X-Axis constitute a horizontal ion beam scan plane. An ion beam is delivered (desirably) parallel to the Z-Axis and strikes the planar surface of the wafer. The X-Axis is horizontally perpendicular to the Z-Axis. The ion beam is scanned back and forth along a scan path parallel to the X-Axis. The Y-Axis is vertically perpendicular to the ion beam scan plane (i.e., the XZ-coordinate plane). The wafer is scanned up and down along another scan path parallel to the Y-Axis by moving the wafer up and down. The depth at which impurities are implanted depends in part upon the angle of incidence of the ion beam along a desired direction, typically perpendicular, to the crystal structure of the semiconductor. Therefore, it is important to control the angle of the ion beam during implantation to maintain a desired direction of the ion trajectories relative to a wafer's crystal structure, particularly when scanning the ion beam across a wafer surface. In particular, in order to achieve repeatable implant results, the angle of the ion beam should be known and controlled to a range of error of less than 1° from parallel to the desired direction, especially for high energy implants and channeled implants. Conventional methods of determining ion beam parallelism are very complex and do not achieve the accuracy of angle determination described above. In addition, conventional methods of correcting ion beam parallelism are incapable of achieving the above-described range of error. There is a need for an improved method of determining parallelism of an ion beam and/or adjusting the ion implanter system based on the parallelism determination. A system, method and program product for determining parallelism of an ion beam using a refraction method, are disclosed. One embodiment includes determining a first test position of the ion beam while not exposing the ion beam to an acceleration/deceleration electrical field, determining a second test position of the ion beam while exposing the ion beam to an acceleration/deceleration electrical field, and determining the parallelism of the ion beam based on the first test position and the second test position. The acceleration/deceleration electrical field acts to refract the ion beam between the two positions when the beam is not parallel, hence magnifying any non-parallelism. The amount of refraction, or lateral shift, can be used to determine the amount of non-parallelism of the ion beam. An ion implanter system and adjustments of the ion implanter system based on the parallelism determination are also disclosed. A first aspect of the invention is directed to a method for determining parallelism of an ion beam in an ion implanter system for implanting into a work piece, the method comprising the steps of: determining a first test position of the ion beam while not exposing the ion beam to an acceleration/deceleration electrical field; determining a second test position of the ion beam while exposing the ion beam to the acceleration/deceleration electrical field; and determining the parallelism of the ion beam based on the first test position and the second test position. A second aspect of the invention is directed to a system for determining a parallelism of an ion beam in an ion implanter system for implanting into a work piece, the method comprising the steps of: a position determinator for determining a first test position of the ion beam while not exposing the ion beam to an acceleration/deceleration electrical field and a second test position of the ion beam while exposing the ion beam to the acceleration/deceleration electrical field; and a parallelism determinator for determining the parallelism of the ion beam based on the first test position and the second test position. A third aspect of the invention is directed to a computer program product for determining a parallelism of an ion beam in an ion implanter system for implanting into a work piece, the computer program product comprising: a computer usable medium having computer usable program code embodied therein, the computer usable medium including: program code configured to determine a first test position of the ion beam while not exposing the ion beam to an acceleration/deceleration electrical field; program code configured to determine a second test position of the ion beam while exposing the ion beam to the acceleration/deceleration electrical field; and program code configured to determine the parallelism of the ion beam based on the first test position and the second test position. A fourth aspect of the invention is directed to an ion implanter system for implanting an ion beam into a work piece, the ion implanter system comprising: an ion beam generator; and a system for determining a parallelism of the ion beam in the ion implanter system including: a position determinator for determining a first test position of the ion beam while not exposing the ion beam to an acceleration/deceleration electrical field and determining a second test position of the ion beam while exposing the ion beam to the acceleration/deceleration electrical field; and a parallelism determinator for determining the parallelism of the ion beam based on the first test position and the second test position. The foregoing and other features of the invention will be apparent from the following more particular description of embodiments of the invention. 1. Definitions In the above and following disclosure, the listed words (phrases) are defined as follows: “Parallelism” is the amount of divergence between an ion beam or an ion trajectory and a desired direction of the ion beam, which is typically parallel to a Z-Axis in an ion beam scan plane. “Angle of incidence” (or simply angle) is the direction by which the ion beam impinges on the wafer surface. The wafer surface may or may not lie in the YZ plane. An “ion beam scan plane” is a horizontal XZ-coordinate plane in which an ion beam is delivered parallel to the Z-Axis and scanned back and forth along a scan path parallel to the X-Axis. A “scan path” is a range along an X-Axis within which an ion beam is scanned. A “scan position” is a lateral position along the scan path. A “test position” of an ion beam is a set lateral position along an X-Axis along the scan path at which parallelism is tested, i.e., a particular scan position. The ion beam may be exposed or not exposed to an acceleration/deceleration electrical field at a test position, as will be described below. 2. Ion Implanter System Overview With reference to the accompanying drawings, FIG. 2 shows an illustrative ion implanter system 10, which may be used in the present invention. Implanter system 10 includes an ion beam generator sub-system 2 for generating and transmitting an ion beam 4, through ion beam scanning sub-system 6, to a work piece sub-system 8. Ion beam generator sub-system 2 may include any now known or later developed ion beam generator such as those available from Varian Semiconductor Equipment Associates. Typically, work piece sub-system 8 includes one or more semiconductor work pieces 12 (e.g., wafer) mounted to a platen 14. Characteristics and positioning of platen 14 and, hence, work piece 12, may be controlled by a platen drive assembly 18 that rotates work piece 12, i.e., wafer, relative to ion beam 4, and/or controls a vertical scan position of work piece 12. Ion implanter system 10 may include additional components known to those skilled in the art. For example, work piece sub-system 8 typically includes automated work piece handling equipment for introducing work pieces into ion implanter system 10 and for removing work pieces after implantation, a dose measurement device, an electron flood gun, etc. It is understood that the entire path traversed by ion beam 4 is evacuated during ion implantation. Besides the above-described components, ion beam generator sub-system 2 may include a gas flow delivery 40, an ion beam source 42, an extraction manipulator 44, a filter magnet 46, a first acceleration/deceleration column 48, and a mass analyzer 50. Filter magnet 46 is preferably positioned in close proximity to ion beam source 42, and precedes first acceleration/deceleration column 48. Extraction manipulator 44 is positioned between filter magnet 46 and ion beam source 42. Mass analyzer 50 may include, for example, a dipole analyzing magnet 52, a mass slit 54 having a resolving aperture 56, and an electrostatic lens 58. Scanning sub-system 6 may include, for example, a scanner 60, an angle corrector 62 and a second acceleration/deceleration column 80. Scanner 60, which may be an electrostatic scanner, deflects ion beam 4 to produce a scanned ion beam 4 having ion trajectories which diverge from a scan origin 64. Scanner 60 may include spaced-apart scan plates 66 and 68 responsive to a scan generator 70. Scan generator 70 generates a scan voltage waveform, such as a sawtooth, or triangular waveform, for deflecting ion beam 4 in accordance with the electric field between scan plates 66 and 68. Angle corrector 62 is designed to deflect ions in scanned ion beam 4 to have parallel ion trajectories, i.e., to focus scanned ion beam 4. In one embodiment, angle corrector 62 may include magnetic pole pieces 72 that are spaced apart to define a gap, and a magnetic coil 74 that is coupled to a power supply 76. Scanned ion beam 4 passes through the gap between pole pieces 72 and is deflected in accordance with a magnetic field in the gap. The magnetic field may be adjusted by varying the current through magnetic coil 74 which is provided by power supply 76. A second acceleration/deceleration column 80 is positioned between angle corrector 62 and work piece sub-system 8, and is capable of applying an electrical field to accelerate ion beam 4 or decelerate ion beam 4. In one embodiment, it is within this acceleration/deceleration column 80 that the teachings of the current invention are applied. Ion implanter system 10 may further include an ion implant/ion beam control system 20. Control system 20 includes at least one ion beam profiler 22 and a controller 24. In one embodiment, profiler 22 is positioned after the acceleration/deceleration column 80, ideally at the wafer plane. Profiler 22 is coupled to controller 24 to receive measurement parameters from and communicate measurement results to controller 24. Controller 24 is further coupled to ion beam generator sub-system 2 and ion beam scanning sub-system 6 to set up/adjust system parameters. Specifically, controller 24 is coupled to, inter alia, extraction manipulator 44, filter magnet 46, mass analyzer 50, electrostatic lens 58, scan generator 70, power supply 76, second acceleration/deceleration column 80, etc. Additional features of control system 20 including profiler 22 and controller 24 will be further described in detail below. Referring to FIG. 2 and 3, in one embodiment, profiler 22 includes a traveling faraday cup system 90. Faraday cup system 90 is positioned in a lateral traveling line 92 substantially parallel to the X-Axis (FIG. 1). That is, traveling line 92 is substantially parallel to the scan path (FIG. 1) of ion beam 4. A desired direction of ion beam 4 is parallel to the Z-Axis. Traveling line 92 is located ideally at the work piece plane. Although an illustrative schematic embodiment of profiler 22 has been illustrated above, it should be understood by those skilled in the art that any now known or later developed system or method to measure an ion beam profile may be used in the current invention and is within the scope of the current invention. For example, although the above described embodiment includes a single faraday cup system 90, more than one (n≧1) faraday cup can be used in a profiler of the current invention. Moreover, although the above described embodiment includes a traveling faraday cup system 90, multiple (n≧1) stationary faraday cups can be positioned along traveling line 92 (or parallel to traveling line 92) to replace traveling faraday cup system 90. FIG. 4 shows an illustrative profile of a spot ion beam 4 (FIG. 2). In FIG. 4, a density of ion beam current (A/mm) is shown as a function of a lateral position along X-Axis (FIG. 1). It should be recognized that density is measured in units of length, rather than area, because the measurement is made across a width interval of ion beam 4. As stated above, a lateral position of an ion beam is set along an X-Axis along the scan path at which parallelism is tested. In one embodiment, the test position is defined by the ion beam center position (CP) that corresponds to the horizontal geometry center of the area under the ion beam profile. Ion trajectories represented by the density of ion beam current corresponding to the center position strike a work piece in approximately the local ion beam angle mean of the ion beam 4. An ion beam width (W) indicates a space between two positions that correspond to zero density of ion beam current, which corresponds to the local ion beam angle spread. Alternatively, the position can be defined by either ion beam side position (SP), i.e., where zero density of ion beam current is observed, so long as the position determinations, described below, consistently use the same side position (SP). Although an exemplary ion implanter system 10 (FIG. 2) has been illustrated above, it should be understood by those skilled in the art that any now known or later developed system to generate and scan ion beam 4 (FIG. 2) may be used for the current invention. It is well known in the art how an ion beam 4 is generated by generator sub-system 2 (FIG. 2) and is scanned by scan sub-system 6 (FIG. 2). Therefore, description of those processes is not necessary for the understanding of the current invention. However, it should be understood that the current invention can be used with any now known or later developed process and methods of ion implantation. 3. Controller Overview Referring to FIG. 5 in conjunction with FIG. 2, a block diagram of an illustrative controller 24 is shown. Controller 24 includes a computer control system responsive to, inter alia, ion beam generator sub-system 2, ion beam scanning sub-system 6, work piece sub-system 8 and profiler 22. In one embodiment, controller 24 includes a memory 240, a processing unit (PU) 242, input/output devices (I/O) 244 and a bus 246. A database 248 may also be provided for storage of data relative to processing tasks. In one embodiment, memory 240 includes a program product 250 that, when executed by PU 242, comprises various functional capabilities described in further detail below. Memory 240 (and database 248) may comprise any known type of data storage system and/or transmission media, including magnetic media, optical media, random access memory (RAM), read only memory (ROM), a data object, etc. Moreover, memory 240 (and database 248) may reside at a single physical location comprising one or more types of data storage, or be distributed across a plurality of physical systems. PU 242 may likewise comprise a single processing unit, or a plurality of processing units distributed across one or more locations. I/O 244 may comprise any known type of input/output device including a network system, modem, keyboard, mouse, scanner, voice recognition system, CRT, printer, disc drives, etc. Additional components, such as cache memory, communication systems, system software, etc., may also be incorporated into controller 24. As shown in FIG. 5, program product 250 may include a parallelism determining system 252 that includes a position determinator 260 including a measurement parameter determinator 262, a parallelism determinator 264 including a lateral shift determinator 266 and a parallelism calculator 268, an adjuster 270 and other system components 272. Other system components 272 may include any now known or later developed parts of an ion implantation controller not individually delineated herein, but understood by those skilled in the art. Referring to both FIGS. 2 and 5, inputs to controller 24 include parameter inputs 280 and measurement inputs 282. Parameter inputs 280 include those from a wide variety of ion implanter system 10 components including, for example, ion generator sub-system 2, ion beam scanning sub-system 6, profiler 22, acceleration/deceleration column 80, and user entered or other parameter inputs. Parameter inputs 280 may indicate, among other things, particular states of ion implanter system 10 including profiler 22 and/or particular components thereof or may indicate user defined input parameters. That is, a parameter input 280 may be any characteristic of ion implanter system 10 including profiler 22, user defined constants or other variables that may affect operation of the system 10 including, in particular to the present invention, parallelism and direction of ion beam 4. Based on the above-described components of profiler 22 (FIGS. 2-3) used to measure a profile of ion beam 4, parameter inputs 280 may include, for example, a desired direction of ion beam 4, positions of sampling faraday cup system 90 including lateral line 92, width of the faraday cup openings, etc. Furthermore, based on the above-described components of ion implanter system 10, parameter inputs 280 may include, for example, filter magnet 46 voltage, source dopant gas 40 flow rate, extraction manipulator 44 positioning (e.g., X, Y, Z axis), mass slit 54 aperture 56 opening, scan plates 66 and 68 spacing, scan generator 70 output (voltage, waveform, direct/alternative, etc), magnetic pole pieces 72 spacing, power supply 76 output, work piece vertical scan system position 18 control setting and/or horizontal (XZ-coordinate plane) ion beam scan speed. Measurement inputs 282 include the results of a measurement by profiler 22. Controller instructions 284 include the instructions to adjust system parameters similar as those received as parameter inputs 280. For example, controller instructions 284 may include instructions to adjust an angle of work piece 12 or an optical component, e.g., corrector magnet 62, electrostatic lens 58, etc., of ion implanter system 10. It should be recognized that the above-described list is meant to be illustrative only. For example, it is common for a conventional controller to receive more than 5000 parameter inputs and create as many controller instructions, depending on the makeup of the ion implanter system used. 4. Parallelism Determineng System Parallelism determining system 252 functions generally to determine parallelism of ion beam 4, i.e., any angle divergence from parallel, using a refraction method. In one embodiment, as mentioned above, the teachings of the invention are applied within second acceleration/deceleration column 80, which is positioned just upstream from work piece sub-system 8. However, it should be recognized that the acceleration/deceleration electrical field may be applied at a different test area within ion implanter system 10. One embodiment of operation of parallelism determining system 252 is shown in the flow diagram of FIG. 6. The illustrative embodiment of operation will be described with reference to FIGS. 2 and 5-7. In step S1, the process starts with controller 24 setting up system parameters of ion beam implanter system 10 in any conventional manner, including, for example, setting up ion beam generator sub-system 2 and scanning sub-system 6 to generate, transmit and scan an ion beam 4 according to the requirements with respect to a specific work piece 12, including a desired direction of ion beam 4. The actual ion beam 4 generated and scanned, however, may not be parallel and may be divergent from a desired direction as it enters work piece sub-system 8. Next, in step S2, position determinator 260 determines a first test position of ion beam 4 while not exposing the ion beam to an acceleration/deceleration electrical field. That is, a first test position of ion beam 4 is determined by position determinator 260 using measurement parameter determinator 262 obtaining a profile and position of ion beam 4 at a first test position, i.e., using profiler 22, while no electrical field is applied by acceleration/deceleration column 80. As illustrated, profiler 22 is after the second acceleration/deceleration column 80. The first test position represents the position of ion beam without being accelerated or decelerated by the acceleration/deceleration electrical field of second acceleration/deceleration column 80. In step S3, position determinator 260 determines a second test position of ion beam 4 while exposing ion beam 4 to an acceleration/deceleration electrical field. That is, a second test position of ion beam 4 is determined by position determinator 260 using measurement parameter determinator 262 obtaining a profile and position of ion beam 4 at a second test position, i.e., using profiler 22, while an electrical field to either accelerate or decelerate ion beam 4 is applied by second acceleration/deceleration column 80. The second test position represents the position of ion beam 4 refracted by the electrical field. Any divergent angle from parallel exhibited by ion beam 4 is therefore magnified by the electrical field. Referring to FIG. 7, a two-dimensional graph showing a calculation of the beam path for first test position 300 (no electrical field) of an ion beam 4 and second test position 302 (with electrical field) of an ion beam 4E is shown. Each test position 300, 302 was calculated for a particular scan position of ion beam 4. In the graph, the horizontal axis represents an actual Z-axis, which extends the longitudinal length of a test area within second acceleration/deceleration column 80, and the vertical axis represents an actual X-axis, which extends laterally across the test area. In this example, the following parameters were used: an extraction voltage for ion beam 4 (energy entering acceleration/deceleration column 80) was approximately 70 kV, an acceleration voltage of approximately 190 kV, a charge state of n=1 and an atomic mass unit (AMU) of approximately 40. As illustrated in the graph, first test position 300 of ion beam 4 is at approximately 8 mm along the X-axis across work piece 12, and second test position 302 of ion beam 4E is at approximately 5 mm along the X-axis across work piece 12. That is, ion beam 4E is laterally shifted (or refracted) (along the X-axis) by the acceleration electrical field by approximately 3 mm. This lateral shift (S) indicates ion beam 4 is divergent from parallel when it enters acceleration/deceleration column 80, and hence would be angled when it encounters work piece 12. If ion beam 4 were not divergent (i.e., substantially parallel) when it enters acceleration/deceleration column 80, the first and second test position 300, 302 would be substantially the same and there would be little or no lateral shift. As also shown in FIG. 7, faraday cup system 90 is positioned at approximately 900 mm from a starting end of acceleration/deceleration column 80. However, faraday cup system 90 may be positioned at any location sufficiently separated from the acceleration/deceleration column 80 to observe the refraction of an ion beam under the influence of the electrical field. Returning to FIGS. 5 and 6, in step S4, parallelism determinator 264 determines a parallelism of ion beam 4 based on first test position 300 and second test position 302. In one embodiment, step S4 includes a number of sub-steps. In sub-step S4A, lateral shift determinator 266 determines a lateral shift (S) between first test position 300 and second test position 302. That is, lateral shift determinator 266 determines the lateral distance between first test position 300 and second test position 302. Next, in sub-step S4B, parallelism calculator 268 determines the parallelism of ion beam based on lateral shift (S). This sub-step S4B can be implemented in a number of ways. In one embodiment, parallelism is calculated based on calibrated empirical data. For example, in one illustrative test, empirical data is used to calculate parallelism for one illustrative species, argon (Ar+), and extraction voltage (energy at entrance to test area), 60 kV. Measurements were made with no acceleration/deceleration electrical field, i.e., 0 kV accel, applied at three test positions: left, center and right. The left test position had a scanner voltage of 5.7 kV and a position of 55 mm, the center test position had a scanner voltage of 11.5 kV and a position of 155 mm, and the right test position had a scanner voltage of 17.2 kV and a position of 255 mm. A corrector magnet current (K lens) for all test positions was 73.8 A, and a target current ranged from 47-66 μA. Measurements were also made with exposure to acceleration electrical field of 100 kV at the corresponding test positions: left, center and right, respectively. Results indicated a lateral shift for the left and center test positions of −1 mm lateral shift, which indicates the ion beam is not parallel. Similar data can be obtained using a deceleration electrical field of, for example, −30 kV. Based on the above data, a calibration algorithm can be implemented indicating parallelism or an angle of an ion beam. The data described above, for example, indicates an angle of 0.2°. Step S5 represents an optional step in which the first and second test position determining step (S4A) is repeated for a plurality of scan positions of ion beam 4. For example, three scan positions may be used as test positions: left, center and right. In other words, ion beam 4 is moved to various scan positions at which it is expected to be used and faraday cup system 90 is used to determine the first and second test position for each scan position. In this case, the parallelism determining step (S4B) includes determining parallelism based on each of the first and second test positions. In another optional step S6, adjuster 270 adjusts ion implanter system 10 based on a result of the parallelism determining step S4. In one embodiment, this step includes adjusting at least one of the following: an angle of the work piece and an optical component of ion implanter system 10. The angle of work piece 12 can be adjusted using platen drive assembly 18 to accommodate the parallelism of ion beam 4. The optical component may include any part of ion implanter system 10 capable of adjusting an angle of ion beam 4, e.g., corrector magnet 42, scanner 60, electrostatic lens 58, etc. 5. Conclusion In the previous discussion, it is understood that the method steps discussed may be performed by a processor, such as PU 242 of controller 24, executing instructions of program product 250 stored in a memory. It is understood that the various devices, modules, mechanisms and systems described herein may be realized in hardware, software, or a combination of hardware and software, and may be compartmentalized other than as shown. They may be implemented by any type of computer system or other apparatus adapted for carrying out the methods described herein. A typical combination of hardware and software could be a general-purpose computer system with a computer program that, when loaded and executed, controls the computer system such that it carries out the methods described herein. Alternatively, a specific use computer, containing specialized hardware for carrying out one or more of the functional tasks of the invention could be utilized. The present invention can also be embedded in a computer program product, which comprises all the features enabling the implementation of the methods and functions described herein, and which—when loaded in a computer system—is able to carry out these methods and functions. Computer program, software program, program, program product, or software, in the present context mean any expression, in any language, code or notation, of a set of instructions intended to cause a system having an information processing capability to perform a particular function either directly or after the following: (a) conversion to another language, code or notation; and/or (b) reproduction in a different material form. While shown and described herein as a method, system and computer product for determining ion beam parallelism and direction integrity, it is understood that the invention further provides various alternative embodiments. For example, in another embodiment, the invention provides a business method that performs the process steps of the invention on a subscription, advertising, and/or fee basis. That is, a service provider, such as an Application Service Provider, could offer to determine ion beam parallelism and direction integrity, as described above. In this case, the service provider can create, maintain, and support, etc., a computer infrastructure, such as a controller 24 (FIG. 5) that performs the process steps of the invention for one or more customers. In return, the service provider can receive payment from the customer(s) under a subscription and/or fee agreement and/or the service provider can receive payment from the sale of advertising space to one or more third parties. In still another embodiment, the invention provides a method of generating a system for determining ion beam parallelism and direction integrity. In this case, a computer infrastructure, such as controller 24 (FIG. 5), can be obtained (e.g., created, maintained, having made available to, etc.) and one or more systems for performing the process steps of the invention can be obtained (e.g., created, purchased, used, modified, etc.) and deployed to the computer infrastructure. To this extent, the deployment of each system can comprise one or more of (1) installing program code on a computing device, such as controller 24 (FIG. 5), from a computer-readable medium; (2) adding one or more computing devices to the computer infrastructure; and (3) incorporating and/or modifying one or more existing systems of the computer infrastructure, to enable the computer infrastructure to perform the process steps of the invention. The foregoing description of various aspects of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously, many modifications and variations are possible. Such modifications and variations that may be apparent to a person skilled in the art are intended to be included within the scope of the invention as defined by the accompanying claims. |
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description | The present invention relates to x-ray optical systems. Researchers have long employed focusing x-ray optics in x-ray diffraction experiments to increase the flux incident on a sample and to thereby increase the signal to noise ratio. A focusing optic increases the flux through a sample by directing a large number of photons from a source. Moreover, by positioning a detector near or at the focus of the optic, resolution of the system can be greatly improved. The intensity of conventional x-ray beam systems, however, is limited by the brilliance of the source that can be achieved without damaging the source target. Although a large optic, such as an ellipsoidal optic or a parabolic optic with a large capture angle, can deliver high flux, the cross section of the x-ray beam and divergence tends to be too large to be fully utilized. Improving the performance of an x-ray beam system by increasing the brilliance of the source is typically limited or too costly. In view of the above there is a need for an improved x-ray optical system that produces high-intensity x-ray beams. In satisfying the above need, as well as overcoming the enumerated drawbacks and other limitations of the related art, the present invention provides an x-ray optical system for producing high intensity x-ray beams. The system includes an optic with a surface formed by revolving a defined contour around a revolving axis that is different than the geometric symmetric axis of the optic and a source that has a circular emission profile. The axis can be a beam axis, the cord of geometric contour of the optic, or a line near the cord of the geometric contour. The optic can be a total reflection mirror or a reflector with performance enhancement coating such as a multilayer coating. The source can be a rotating anode or a sealed tube x-ray generator. Further features and advantages will be apparent from the following description and from the claims. The invention provides x-ray optical system that produces high intensity x-ray beams. In accordance with an embodiment of the invention, FIG. 1 illustrates an x-ray optical system 10 including an x-ray source 12 and an optic 14. The x-ray source 12 can be a laboratory source, such as a high brilliance rotating anode, a sealed tube x-ray generator, or a microfocusing source. The optic 14 can be a reflector with a performance enhancement coating, such as a multilayer coating, a total reflection optic, or an x-ray reflective crystal. In operation, the source 12 emits an x-ray beam 16 toward the optic 14. The optic 14 includes an optical surface 26 that directs the x-ray beam 16 onto a sample S, for example by focusing or collimating the x-ray beam 16. The optic 14 may serve to converge the x-ray beam 16 as it is directed onto the sample S thereby increasing the amount of flux provided to the sample S. The detector 18 may collect radiation 22 from the x-ray beam 16 that is transmitted and diffracted through the sample S. The detector 18 may provide a signal 24 to a processor 20 corresponding to the radiation 22 collected by the detector 18. The optic 14 may have various geometric contours along the optical surface 26 in the direction of x-ray beam propagation. Such contours may include elliptical, parabolic, and hyperbolic surfaces, although other surfaces may also be used. The optical surface 26 is formed by revolving a defined contour around a revolving axis 30. Typically, the contour is revolved around the geometric symmetric axis of the contour. However, in the embodiment described, the optical surface 26 is formed by rotating the contour around a revolving axis 30 that is different than the geometric symmetric axis of the contour. The revolving axis 30 is chosen to provide a beam with a small cross section. Such an axis can be the beam axis 32, a cord of the geometric contour, or a line near the cord of the geometric contour. In addition, the revolving axis may be in plane with the geometric axis of the contour. Since the optical surface 26 will be revolved about the revolving axis 30, the profile of the optical surface 26 will be circular in a cross-section perpendicular to the revolving axis. As such, the cross section will not be perpendicular to the geometric axis of the surface. The optic 14 may have various geometric profiles longitudinally along the length of the optic 14, depending on the requirements of the particular application. In certain implementations, the optic 14 is a semi-circular concave multilayer focusing/collimating optic and the source 12 has a circular emission profile. In other implementations, the optic 14 is a full circular convex multilayer focusing/collimating optic. In yet other implementations, the optic surface is a total reflection surface produced, for example, by controlled pulling from a glass tubing or by replicating technology. A particular feature of the system 10 is that it provides increased flux while maintaining the same beam divergence or convergence. As described above, optical surface 26 of the optic 14 is a 2D surface. This may also be referred to a 3D surface meaning the surface is in 3-dimensional space. The optical surface 26 can be described by its curvature in the “longitudinal” direction 28 and in the “cross” direction (which is perpendicular to FIG. 1 and shown as 34 in FIG. 2). The optical surface 26 along the longitudinal direction 22 can be any geometry surface such as elliptical, parabolic, or hyperbolic curves. The optical surface 26 in the cross direction 34 is formed by revolving the longitudinal curve about a revolving axis 30. In conventional systems, if the revolving axis 30 is defined by two focuses of an elliptical curve, it is an ellipsoidal surface, or if the axis 30 is defined by the symmetric axis of a parabola, it is a parabolic surface. However, the surface 26 is formed by revolving the longitudinal curve (or contour) around an axis 30 that is equal or close to the intended beam axis 32. Thus, rather than producing a large ring with conventional ellipsoidal/paraboloidal surfaces, the optic 14 delivers a small ring, a semi-circular ring, or a round beam if the revolving axis 30 is the beam axis 32. The specific shape of the x-ray beam 16 depends on the relative relationship between the revolving axis 30 and the contour. The optic 14 can have a concave surface, a convex surface, or a combination of a concave surface and a convex surface. If the optic 14 is concave surface, the surface many not be a fully closed surface (depending on where the revolving axis is). If the optic 14 is convex, a 2π surface can be formed. The source 12 may have a circular or a partial circular profile. The diameter of the partial-circular source (or full circular source) profile depends on the optic design. If the optic is a combination of convex and concave, one part can be a 2π surface while the other part may be a semi-circular π surface. A system 310 utilizing an optic 314 with an elliptical surface 326 is provided in FIG. 3. In an elliptical system, the origin of the x-ray source 312 is located at one focus 340 of the ellipse 344 while the detector is generally located at or near the second focus 342 of the ellipse 344. The system has a revolving axis 330 different from the geometric axis 350. The revolving axis 330 passes the second focal point Typical systems use an optic with a surface that is rotated about the geometric symmetric axis 350 of the contour. (The revolving axis 330 is the same as the geometric axis 350). In the case of an elliptical surface 326, the geometric symmetric axis 350 passes between the two focuses 340, 342. This geometry works particularly well with an x-ray source that emulates a point source. However, point sources are limited in the amount of power per unit area that can be generated. This is further illustrated in FIG. 4. A typical x-ray point source 410 generates a line 412 on a target 414 that is then viewed at a very shallow angle 418 that may appear as a point source 416. However, to increase the brilliance of the source, more energy must be projected into the area. Since the tolerable heat generated by the electrons projected on the target 414 is limited, it is useful to generate an optical system that can use a large area than a point source. For example, an x-ray source with a circular emission profile may be used to maintain a brilliance not much lower than a point source, but can be loaded with much higher power due to the larger area. As shown in FIG. 5, a circular x-ray source may be developed by projecting accelerated electrons 510 onto a rotating anode target 512 with a semi-circular profile 514. In this scenario, the x-ray source and the optic will need to be designed as a matching pair based on the particular configuration. One embodiment, may include a partial-circular concave multilayer optic and a source with a partial circular emission profile. Another embodiment may include a full circular convex/concave multilayer focusing optic and a source with a circular emission profile. Yet another embodiment, may include either geometry of the sources or optic but utilizing a total reflection surface, which may be formed by controlled pulling from a glass tubing. It is also valid that such an optic would be able to take advantage of a large source provided that source size is large enough the perceived circular source is included within the large source. Referring now to FIG. 6, there is shown a schematic relationship between the revolving axis and the resulting reflecting surface 602. In each case shown, the axis of rotation may pass through the focus 606 and be aligned in plane with the geometric symmetric axis of the contour. If the axis of rotation is beyond the far edge of the reflecting surface 602 as shown by revolving axis 610, the reflecting surface 602 is a convex surface, and the cross section of the beam formed has a ring with a center hole. If the revolving axis rotation is at the far edge of the reflecting surface 602 as shown by revolving axis 612, the reflecting surface 602 is a convex surface, and the beam formed has a partial or full round cross section, depending on the revolving angle. The reflecting surface 602 may be a combination of a concave surface and a convex surface, if the axis of rotation is between the near edge of the reflecting surface 602 and the far edge of the reflecting surface 602 as shown by revolving axis 614. The concave surface is at most a π surface. The cross section of the beam is round if the revolving angle is π. An illustration of the optic and x-ray beam projection is provided in FIG. 7. The optic 710 has a convex portion 712 and a concave portion 714. To utilize the convex portion 712 of the optic 710, the x-ray source 716 must be a partial circle. This allows more flux to be directed to the sample without concentrating more energy into a small area of target on the x-ray source 716. A trace of the x-ray beam is denoted by reference numeral 718. If the axis of rotation is at the near edge of the reflecting surface 602 as shown by axis of rotation 616, the reflecting surface 602 is a concave surface. The revolving angle is at most π. As such, the beam has a semi-circular cross section. If the axis of rotation is between the source 604 and the near edge of the reflecting surface 602 as shown by axis 618, the reflecting surface is a concave surface. The beam has a cross section of a partial ring with an inner diameter that is smaller than that provided by the typical case shown by revolving axis 620. If the revolving axis 620 passes through the origin of the source 604 and the focus 606 as the reflecting surface is an ellipsoidal surface. The beam has a cross section of a ring or a part of a ring. A system 810 utilizing a parabolic surface is provided in FIG. 8. In a parabolic system, the origin of the x-ray source 812 is located at the origin of the parabola 844. Typical parabolic systems use an optic 814 with a surface 826 that is revolved about the geometric axis 850 of the surface 826. However, in the system according to this embodiment of the invention, the revolving axis 830 is different than the geometric symmetric axis 850 of the surface 826. As discussed above, this serves to substantially increase the flux delivered to the sample in an efficient and effective manner if the axis 830 is between the axis 850 and the optic. Axis 830 is in parallel to axis 850. In x-ray analysis, being able to deliver multiple wavelengths or energies onto a sample may be useful. As such, the x-ray source 12 may be optionally configured to deliver a beam of multiple wavelengths or multiple energy. Alternatively, the circular shaped source may be further segmented into multiple sections of different target materials, as shown in FIG. 9. For example, the target 910 has four sections 912, 914, 916, and 918 each made of a different material. Such a source-optic combination would deliver a beam of multiple wavelengths. The optic 920 of a multiple wavelength system may be a total reflection optic in nature or a diffraction optic in nature, such as multilayer optics or crystal optics. If the reflection surface is in nature of the multilayer or crystal surfaces, the corresponding sections 922, 924, 926, and 928 for different energies will follow their Bragg's law governed contours and layer structures. In one example, the optic includes corresponding sections for different energies and each section follows Bragg's law with its own contour and coating structure which include layer thickness and variation of the layer thickness. In another example, each section has the same contour for each section but different coating structure for each section. In yet another example, each section of the optic has the same coating structure but different contours. In again another alternative example, the optic is a crystal optic with different sections, each of them has its own contour and crystal structure so that Bragg's law can be satisfied for its energy. As a person skilled in the art will readily appreciate, the above description is meant as an illustration of implementation of the principles this invention. This description is not intended to limit the scope or application of this invention in that the invention is susceptible to modification, variation and change, without departing from the spirit of this invention, as defined in the following claims. |
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051732483 | claims | 1. A remote control apparatus for maintaining a tokamak type nuclear fusion reactor comprising a vessel having a torus space formed therein and a plurality of maintenance ports extending radially for communicating said torus space with the outside of said vessel, and in-vessel components arranged in said vessel, comprising: a rail having generally similarly arcuated links pivoted to one after another and extendable in the circumferential directions in said torus space, said links forming, in said torus space, a continuous semi-circular arc having a center substantially coincident with a center of said torus space when said links are extended; a vehicle carrying at least one manipulator for handling said in-vessel components and guided along said rail extended in said torus space; rail housing means arranged outside of said vessel, for housing said rail when said remote control device is not used; rail mounting means supporting the last link of said rail and delivering, into said torus space, said links housed in said rail housing means from the first link to the last link in succession through one of the maintenance ports and causing the links to form the continuous semi-circular arc; and a rail supporting device inserted in said torus space through another maintenance port adjacent to the first mentioned maintenance port, for supporting a central portion of said rail. controlling means for controlling said mechanisms in position, speed and force; monitoring means for monitoring said mechanisms in operating force and position, and for signaling a measured data to monitor the operating condition of said mechanisms; performing changeover means for changing the operating mode of said controlling means to change the rigidity of the position control for said mechanisms depending on the monitored operating condition of said mechanisms; and generating command means for generating commands to correct said mechanisms in position, speed and force depending on the monitored operating condition of said mechanisms. 2. An apparatus according to claim 1, wherein said rail comprises a plurality of joints for preventing the adjacent arcuated links from rotating inwardly thereof and allowing said adjacent arcuated links to rotate outwardly, and said rail housing means guides said arcuated links in a folded state on one after another by rotating said arcuated links outwardly around the respective joints. 3. An apparatus according to claim 1, wherein said rail mounting means comprises a fixed guide fixed at the outside of said torus space at a position on a line extended from said one maintenance port, slide links slidable in said one maintenance port and a rail carriage movable along said slide links for sending said arcuated links in succession from said rail housing means into said torus space through said one maintenance port. 4. An apparatus according to claim 3, wherein one of said slide links is a second link driven along said fixed guide and the other link is a first link driven along said second link and having guide means for sending said arcuated links in said one maintenance port. 5. An apparatus according to claim 4, wherein said first link has an end close to said torus space and a support arm drivingly pivoted to said end, said rail carriage has fixed thereto a swing link pivoted by means of one of said joints to the arcuated link which is sent to said torus space last, for supporting said rail via said joint which is sent last when said rail is extended to form a continuous arc in said torus space. 6. An apparatus according to claim 5, wherein said first slide link has a guide groove for housing rollers and a roller holder driven along said guide groove, for causing said rollers of said rail to slide in said guide groove for a predetermined stroke (A). 7. An apparatus according to claim 5, wherein said vehicle is detachably mounted on said support arm and has driving means for moving said arcuated links along said arc defined by the extended arcuated links. 8. An apparatus according to claim 5, wherein said swing link is intermittently driven for sending said arcuated links from said rail housing means to said first slide link, said roller holder reciprocates in said guide groove in synchronism with said swing link, and said driving means for moving said vehicle are driven in synchronism with said swing link and said roller holder. 9. An apparatus according to claim 1, wherein said rail has a plurality of mechanical locking mechanisms for locking the adjacent arcuated links so as not to be pivoted when said adjacent links form an arc, and said vehicle has operating means for actuating said mechanical locking mechanisms. 10. An apparatus according to claim 9, wherein each of said mechanical locking mechanisms comprises a hook having a support shaft pivoted to an end of one of the adjacent two arcuated links, said support shaft being provided with a projection extending toward an inner side of said one arcuated link, and said operating means has a recess engaging with said projection. 11. An apparatus according to claim 1, wherein said rail has a projection formed on an outer periphery of the arcuated link adjacent to said another maintenance port, and said rail supporting device comprises a base disposed at the outside of said torus space, a slide link reciprocating along said base in said another maintenance port and having a distal end which is accessible to said rail when said slide link is advanced, and a pair of holding members provided on said distal end of said slide link, for firmly holding said projection of said rail. 12. An apparatus according to claim 11, wherein said distal end of said slide link has a guide roller for supporting an undersurface of the respective arcuated link of said rail moved in an circumferential direction of said torus space in said torus space when said rail is extended in said torus space, said guide roller being selectively moved to a position at which said roller supports said respective arcuated link and to a position remote from said respective arcuated link. 13. An apparatus according to claim 1, further comprising an in-vessel component transporting device disposed below said rail supporting device, for transporting said in-vessel components from an interior of said torus space to the outside thereof and from the outside of said torus space to said interior thereof through said another maintenance port. 14. An apparatus according to claim 1, wherein said vehicle is detachably provided with another manipulator and a transporting device transports said another manipulator. 15. An apparatus according to claim 1, wherein said vehicle is movable over the whole length of said rail extended in said torus space, and comprises a frame having an inner passage, an opening for preventing interference with said joints of said rail and said projection adapted to be held by the paired holding members and a plurality of wheels rolling on an outer surface of said rail. 16. An apparatus according to claim 15, wherein said vehicle driving means has a servo motor and a pinion driven by said servo motor, and said arcuated links are provided in an inner surface thereof with a rack engageable with said pinion. 17. An apparatus according to claim 1, further comprising a cable system comprising cable means connected to said vehicle, for supplying a power to said vehicle from the outside thereof and transmitting signals between said vehicle and an external control device, a cable drum disposed at the outside of said torus space, for taking up and taking off said cable means, a drive roller rotatably mounted on said rail carriage, for moving said cable means in a direction according to a position of said swing link, and a plurality of cable supports mounted on the respective joints, for taking up and taking off said cable means. 18. An apparatus according to claim 3, wherein said swing link firmly supports a distal end portion of an arcuated link at a distal end of another rail extended through said another maintenance port after the last arcuated link has been sent into said torus space so as to form a circular rail assembly together with the first mentioned rail in said torus space. 19. An apparatus according to claim 18, wherein said vehicle is moved along said first mentioned rail and said another rail. 20. An apparatus according to claim 1, further comprising a plurality of driving mechanisms with servo motors at least for extending and restoring the rail and for handling said in-vessel components, and a control device for said plurality of driving mechanisms, said control device including: 21. An apparatus according to claim 1, wherein said vehicle carries a manipulator having a telescopic arm and an actuator for swinging said telescopic arm in a plane perpendicular to said rail. 22. An apparatus according to claim 2, further comprising rotatable rollers coaxially provided under said joints, said rail housing means having a guide groove for guiding said rollers, and said guide groove being formed in a circular form. |
042960746 | claims | 1. A method of decladding an assembly comprising an element selected from the group consisting of uranium, thorium and mixtures thereof, clad in stainless steel, zirconium or a zirconium alloy consisting essentially of zirconium and containing minor amounts of nickel, chromium, tin, iron and combinations thereof; said method comprising: perforating the cladding material to expose the selected element; reacting said selected element with hydrogen at a temperature of from 450.degree. C. to 680.degree. C. and a hydrogen pressure of from 360 to 1400 torr for a time sufficient to form a hydride of substantially all of the selected element; increasing the temperature to a range of from about 700.degree. C. to 900.degree. C. to dehydride the selected element to reduce it back to its elemental form; separating the cladding material from the selected element; and recovering the selected element in a friable particulate readily comminutable form. 2. The method of claim 1 wherein the selected element is hydrided and dehydrided from two to four times. 3. The method of claim 1 wherein the selected element is hydrided and dehydrided at least three times. 4. The method of claim 1 wherein in the hydriding step the selected element is thermally cycled between the temperatures of from 450.degree. C. to 680.degree. C. 5. The method of claim 1 wherein the selected element is reacted with hydrogen at a temperature within the range of from about 500.degree. C. to 650.degree. C. and at a hydrogen pressure of about 760 torr. 6. The method of claim 1 wherein the selected element is a mixture of thorium and uranium. |
claims | 1. A process for monitoring at least one operating parameter of a nuclear reactor core of a nuclear power station power unit, the core having a number of juxtaposed fuel assemblies arranged over the height of the core, the process comprising the steps: introducing a plurality of detectors, each having a string of spaced stacked detector units, into at least some of the fuel assemblies of the core for measuring neutron flux, the detectors being fixed and distributed over the height of the core; during operation of the nuclear reactor, at specified time intervals: a) measuring the neutron bulk flux distribution using a maximum number of detectors being equal to 15% of the number of fuel assemblies; b) employing a neutron flux calculation code in conjunction with measurements provided by the detectors to obtain an instantaneous neutron bulk flux distribution throughout the entire core in the form of a set of neutron flux values at points distributed throughout the core, c) calculating at least one core operating parameter from the instantaneous neutron bulk flux distribution; and d) raising an alarm if at least one operating parameter is outside a preselected range. 2. The monitoring process according to claim 1 , together with the following steps to obtain the instantaneous neutron flux distribution throughout the entire core: claim 1 instantaneously calculating, on the reactor site, the bulk flux distribution inside the core, in the form of a set of neutron flux values at the various points distributed throughout the core, comprising a first sub-set of instrumented positions where the neutron flux measuring detectors are located and a second sub-set of non-instrumented positions, the instantaneously calculating being based on parameters originating from the power unit plant, using the neutron flux calculation code; calculating, for each instrumented position, the difference between the flux values obtained by measurement and the corresponding values calculated from the parameters originating from the power unit plant; extrapolating the corresponding differences for every non-instrumented position, from the differences relating to the instrumented positions; algebraically adding the values of the instrumented and non-instrumented differences to the bulk flux distribution values obtained from the parameters originating from the power unit plant, to obtain the measured value of the bulk flux distribution for every point distributed throughout the core. 3. The monitoring process according to claim 1 together with the following steps to obtain the instantaneous neutron flux distribution throughout the core: claim 1 instantaneously calculating, on the reactor site, the neutron bulk flux distribution inside the core, in the form of a set of neutron flux values at the various points distributed throughout the core, comprising a set of instrumented positions where the neutron flux measurement detectors are located, the instantaneous calculating being based on parameters originating from the power unit plant, using the neutron flux calculation code; calculating, for each instrumented position, the differences between the neutron flux values obtained by measurement and the corresponding values calculated from the parameters originating from the power unit plant; the calculated differences being used to correct defining parameters of the neutron flux calculation code; and performing a second instantaneous calculation on the nuclear reactor site of the instantaneous neutron flux distribution inside the core based on the parameters originating from the power unit plant, using the neutron flux calculation code which includes corrected defining parameters. 4. The monitoring process according to claim 1 wherein the instantaneous values of the parameters coming from the power unit plant and current neutron flux values determined for every point in the core are used to calculate new instantaneous neutron flux values simply, using the calculation code. claim 1 5. The monitoring process according to claim 1 wherein, for a core containing a number of fuel assemblies approximating 200, fewer than 30 detectors are used. claim 1 6. The monitoring process according to claim 5 wherein, for a core containing 193 fuel assemblies, 16 detectors are used. claim 5 7. The monitoring process according to claim 1 wherein, the at least one operating parameter of the nuclear reactor core is chosen from the group of the following parameters: linear power density P lin , critical heating ratio (CHR, or DNB ratio), axial power imbalance PI ax , azimuthal power imbalance PI az , and negative reactivity margin NRM. claim 1 8. The monitoring process according to claim 1 , wherein conditioning processing of the measured signals from the detectors includes the steps of isolating, for each signal, an electric signal due to a phenomenon of fast electron production; and by applying an inverse transfer function, determining a neutron flux value from electric current isolated from the current signal. claim 1 9. The monitor process according to claim 1 , wherein neutron flux measurements are obtained by reading outputs of the flux detectors at time intervals of less than one minute. claim 1 10. An apparatus for monitoring at least one operating parameter of a nuclear reactor core of a nuclear power station power unit, the core having a number of juxtaposed fuel assemblies arranged over the height of the core, the apparatus comprising: a plurality of detectors, each having a string of spaced stacked detector units, located in at least some of the fuel assemblies of the core for measuring neutron flux, the detectors being fixed and distributed over the height of the core; the maximum number of detectors being equal to 15% of the number of fuel assemblies; means for conditioning signals from the detectors; processing means for determining the power distribution in the core and the at least one core operating parameter; means for comparing the at least one operating parameter with at least one predetermined limit and for generating an alarm signal if the limit is exceeded; and means for displaying the alarm signal. 11. The apparatus according to claim 10 , wherein the neutron flux measurement detectors are self-powered neutron detectors. claim 10 12. The apparatus according to claim 11 , wherein the self-powered neutron detectors comprise a transmitter made of a rhodium-based material. claim 11 |
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abstract | Nuclear reactor fuel assembly comprising fuel elements installed in a frame having guide channels and spacer grids; a bottom nozzle; and a removable head. The head comprising collet tubes, an upper shell, a support element in the form of a tube, and springs. The collet tubes comprise two coaxially arranged tubes that are movable relative to each other and that each have stops on their side surfaces. The stops interact with each other to select the length of the collet tubes. The upper shell has a tube with a rigidly fixed plate interacting with the springs. The plate has plural holes having a shape corresponding to a shape of a respective boss of the support element. The clearance in plan view between a respective hole and a respective boss being at least the mounting clearance between the tube of the support element and the tube of the upper shell. |
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claims | 1. A container for confining spent nuclear fuel, comprising:a sleeve of longitudinal axis comprising a first longitudinal end closed by a base and a second longitudinal free end via which the container is configured to be loaded;a plug configured to close the second longitudinal free end tightly by a weld made at a level of a welding zone, the welding zone being offset radially towards an interior of the container, relative to an inner face of the sleeve; anda flange projecting from the inner face of the sleeve to a side of the second longitudinal free end, the plug having at least one external diameter substantially equal to at least one internal diameter of the flange, the plug configured to be welded on the flange. 2. The container as claimed in claim 1, in which the flange is welded on the inner face of the sleeve. 3. The container as claimed in claim 2, in which the flange comprises a first section of larger internal diameter and a second section of smaller internal diameter connected by an annular surface forming a shoulder substantially orthogonal to the longitudinal axis, the plug comprising an external profile corresponding to the internal profile of the flange, such that the plug rests on the shoulder. 4. The container as claimed in claim 2, further comprising a handling device of the container distinct from the plug, the handling device being fixed to the sleeve at a level of the longitudinal free end of the sleeve, the flange being fixed in the sleeve back from the longitudinal free end of the sleeve. 5. The container as claimed in claim 4, in which the handling device includes a cover fitted in its center with a gripping element. 6. The container as claimed in claim 5, in which the cover comprises a rim of thickness close to the thickness of the longitudinal free end of the sleeve so as to enable welding of the cover on the sleeve without adding wire. 7. The container as claimed in claim 1, further comprising a wedging device of loading, the wedging device being mounted in the container. 8. The container as claimed in claim 7, the container being circular in cross-section, the wedging device comprising a casing whereof a length is substantially equal to that of the interior of the container and whereof transversal cross-section is a square, length of diagonals of the square being substantially equal to the internal diameter of the container. 9. The container as claimed in claim 8, in which the casing comprises a heat sink of heat emitted by loading. 10. The container as claimed in claim 9, in which the heat sink comprises fins extending longitudinally at least on one part of outer faces of the casing in a direction of largest dimension. 11. The container as claimed in claim 1, in which the plug comprises a first and a second connector for draining the container, sweeping with air of the interior of the container and its pressurizing. 12. The container as claimed in claim 11, further comprising a discharge pipe arranged inside the container, the pipe comprising a first free end arranged in the base of the container and a second end connected to one of connectors for suctioning water contained in the container. 13. A process for loading and closing a container for confining spent nuclear fuel including:a sleeve of longitudinal axis including a first longitudinal end closed by a base and a second longitudinal free end via which the container is configured to be loaded,a plug configured to close the second longitudinal free end tightly by a weld made at a level of a welding zone, the welding zone being offset radially towards an interior of the container, relative to an inner face of the sleeve,a flange projecting from the inner face of the sleeve to a side of the second longitudinal free end, the plug having at least one external diameter substantially equal to at least one internal diameter of the flange, the plug configured to be welded on the flange,the process comprising:placing the loading inside the container;placing the plug in the flange;welding the plug on the flange without adding material. 14. A process for loading and closing, as claimed in claim 13, further comprising a subsequent placing a handling device on the longitudinal free end of the sleeve and welding the handling device on the sleeve. |
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052971775 | description | DETAILED DESCRIPTION OF THE INVENTION (1) Deformation of the zirconium alloy members used in a channel box depends on the <0001> crystallographic orientation of hexagonal crystal Zr in the direction substantially normal to the member surfaces. The hexagonal crystal lattice shrinks in the <0001> crystallographic orientation and expands in the direction normal to the <0001> crystallographic orientation, when irradiated with neutrons. More strictly speaking, a dislocated face (atomic face) is introduced in the <0001> crystallographic orientation due to the neutron irradiation, and shrinkage and expansion take place in the above-mentioned specific directions. As a result, elongation takes place in the longitudinal and lateral directions of the channel box, and shrinking deformation takes place in the normal-to-plate direction. Fuel rods elongate in the longitudinal direction. Neutron irradiation exposure is higher towards the center of the reactor core, and is lower towards the periphery of the reactor core. The channel box is provided at the periphery region of the reactor core, where the neutron irradiation dosage drastically changes, and thus there is a difference in the elongation between the surface side facing the reactor core center and the opposite surface side thereto of the channel box. Thus the channel box undergoes curving. The above-mentioned deformation caused by the neutron irradiation does not give rise to any volume change, and even if individual crystal grains of a polycrystalline material undergo deformation in specific directions, respectively, no deformation takes place on the whole, so long as the specific directions are in a random distribution. To suppress the irradiation growth and curving, it is effective to bring the crystallographic orientations into a random distribution. In the present invention, the crystallographic orientations of the channel box and water rod are brought into a random distribution, whereas that of the fuel cladding tubes are not brought into a random distribution. Consequently, only the fuel rods elongate due to the irradiation growth, whereas neither the channel box nor the water rod undergoes elongation. Since the water rod fixed to the upper tie plate undergoes no change in length, the position of the upper tie plate undergoes no change, either. Since the upper end plugs of the fuel rods are not fixed to the upper tie plate, but only pass through the through-holes provided through the upper tie plate, no such force as to push the upper tie plate upwards is generated even if the fuel rods elongate, and the channel box fixed to the upper tie plate is never pushed upwards at all. As a result, the length of fitting allowance is never decreased. That is, the problem of fuel rod elongation can be solved by protruding the upper end plugs of the fuel rods from the upper tie plate through through-holes provided through the upper tie plate. Generally, quantitative evaluation of the orientation parameter of a crystallographic orientation is carried out by measuring the reflectance on a specific crystal face and the refraction intensity of transmitted X-rays and calculating, an F value from the following equation (1): EQU F=.delta..sub.O.sup.2/.pi. V(.phi.).multidot.cos.sup.2 .phi..multidot.d.phi.(1) In the equation (1), .phi. means an angle of a specific crystallographic orientation (for example, <0001> crystallographic orientation) to a specific direction, for example, a normal-to-plate direction, and V(.phi.) is a volume ratio of crystal oriented in the .phi. direction. When crystallographic orientation parameters in the normal-to-plate direction (direction r), the longitudinal (rolling) direction (direction l), and the lateral direction (direction t), which are perpendicular to one another, are designated as Fr, Fl and Ft, respectively, sum total of Fr, Fl and Ft is equal to unity (1), and when the respective values are 1/3, the crystallographic orientations will be brought in a complete random distribution. The <0001> crystallographic orientation of a plate or tube prepared by cold rolling is in a normal-to-plate (or tube) surface direction (direction r), its Fr value is in a range of 0.6 to 0.7, and its Fl value is in a range of 0.05 to 0.15. A channel box was subjected to irradiation by 3 cycles at the center region of a reactor core and by one cycle at the peripheral region of the reactor core and another channel box was also subjected to irradiation by 4 cycles at the center region of the reactor core to calculate influences of Fr value upon the curving degree of the channel boxes. One cycle consisted of an 18-month operation, and the neutron irradiation dosage was about 2.times.10.sup.22 (n/cm.sup.2) when the fuel assemblies were to be removed from the reactor, and these conditions were according to the most standard shuffling pattern. With increasing residence cycles in the peripheral region of the reactor core, the curving degree further increased. Clearance between the control rod and the channel box was about 3.3 mm wide at the initial period of fuel loading, and a deformation span due to expansion deformation was about 2.2 mm long besides that due to the curving deformation. That is, when the curving degree due to irradiation growth of a channel box having had an experience of residence in the peripheral region of the reactor core reached 1.1 mm, there occurred an interference between the channel box and the control rod. It was found that the Fr value of a channel box having had an experience of residence in the peripheral region of the reactor core must be brought into a more random distribution, that is, to at least 0.25, and the Fr value of a channel box having had no experience of residence in the peripheral region of the reactor core must be at least 0.20 to bring about a more random distribution. It is effective for bringing the crystallographic orientations in a random distribution to heat the zirconium alloy material to a .beta.-phase temperature range (.gtoreq.980.degree. C.) to make .beta.-Zr crystal grains grow, and then cool the material. By the heat treatment, hexagonal .alpha.-Zr crystal grains oriented in a specific direction are transformed to cubic .beta.-Zr crystal grains, which are again transformed to hexagonal .alpha.-Zr crystal grains by cooling. The crystallographic system of the Zr crystal grains after cooling to room temperature is the same hexagonal .alpha.-phase as before the heating, but the crystallographic orientations of the materials having had an experience of transformation to the cubic .beta.-phase are in a random distribution. The higher the heating temperature and the longer the heating time, the higher the degree of randomness. To obtain Fr value .gtoreq.0.20, a heat treatment parameter defined by the following equation must be 0.8 or more by controlling the heating temperature and the heating time: EQU P=(3.5+logt).times.log(T-980) t: heating time (h), PA1 T: heating temperature (.degree.C.) Preferably, P.gtoreq.1.5. At P=0.8, .beta.-Zr crystal grains have an average grain size of 50 .mu.m, and at P.gtoreq.1.5 .beta.-Zr crystal grains have an average grain size of 90 .mu.m or more. An average grain size of not more than 300 .mu.m at the maximum is preferable, and 70 to 130 .mu.m is more preferable. F values of the present channel box and water rod are preferably 0.25 to 0.36 for the Fl value, 0.25 to 0.36 for the Ft value and 0.25 to 0.50 for the Fr value. Particularly preferably, the Fr value is more than the Ft value and the Fl value, and it is most preferable that the Fl value is 0.30 to 0.35, the Ft value 0.30 to 0.35 and the Fr value 0.30 to 0.35. Ideally, all of the F values must be 0.3333. These F values depend on the heating temperature and times. Particularly a treating temperature is practically 980.degree. to 1,350.degree. C. and preferably 1,050.degree. to 1,150.degree. C. Preferable retention time at that temperature is as short as about one second to about one minute. The present process for producing a channel box comprises heating a plate material locally in a .beta.-phase temperature region by an induction coil for a desired residence or retention time while continuously moving the plate material, and forcedly cooling the plate material after the heating. By heating to the .beta.-phase temperature region, the <0001> crystallographic orientation can be brought in a random distribution and a higher corrosion resistance to high temperature, high pressure pure water can be obtained. Cooling is preferably carried out by spraying water onto the heated plate material at a cooling rate of 100.degree. C./sec or more, particularly 150.degree. C./sec or more. Infrared heating and electric furnace heating can be also used as other heating means. For the heating in the .beta.-phase temperature region, the Zr-based alloy plate material must be fixed and constricted by members, such as a mandrel, having a higher coefficient of thermal expansion than that of the Zr-based alloy. Particularly in case of a tubular material, the heating and cooling are carried out preferably by inserting a metallic member as a mandrel into the tubular member without entire contact but with some local contact with the inside surface of the tubular member to decrease heat influence, and fixing both members at both ends to prevent deformation of the tubular member due to the heating and cooling. By providing such a constricting member, the heating and cooling can be readily carried out. Preferable constricting member materials include austenite stainless steels such as SUS 304, 316, 347, etc. After the .beta.-phase heat treatment, annealing is carried out at 500.degree. to 650.degree. C. to uniformly heat the entire member material. It is also preferable to conduct the annealing while constricting the member material with the above-mentioned constricting member to rectify the shape of the tubular member. These heat treatments are carried out in a non-oxidative atmosphere, particularly preferably in Ar. After the final heat treatment the oxide film is removed from the surface by sand blasting and pickling. After the removal of the oxide film, the surface is subjected to an oxidation treatment in an autoclave to form a stable oxide film on the surface, making a final product. The edge parts having screw holes, etc. for the fixing at both ends are removed from the final product. Two open channel-formed (or U-shaped) members for the present channel box are butt welded to each other by plasma welding at the open channel edges to form a square cylinder, and then the welding seams are flattened. For the heat treatment of the square cylinder, a X-shaped contricting member is preferable. The present heat treatment can be carried out in the state of a plate material state, 2 channel-formed members or a welded square cylinder. The above-mentioned temperature and time can be also applied to the production of a water rod, and the above-mentioned heat treatments can be carried out at any stage, that is, from the tube shell-shaped stage after the final hot plastic working to a tube-shaped stage after the final cold plastic working, as in the case of the present cladding tubes. However, when cold plastic working and annealing are carried out after the heat treatment, the random distribution of the crystallographic orientation turns anisotropic, and unless the corrosion resistance is most important, it is most preferable to conduct the heat treatment of the water rod together with the channel box after the final cold plastic working. (2) In order to obtain a high corrosion resistance and a low hydrogen pickup ratio, it is important that the Zr-based alloy contains 1 to 2% by weight of Sn, 0.2 to 0.35% by weight of Fe and 0.03 to 0.16% by weight of Ni with or without Cr. Even if such a highly corrosion-resistant, high Fe--Ni, Zr-based alloy member is used in a BWR circumstance, no nodular corrosion appears and the hydrogen pickup ratio is considerably lower than that of currently available zircaloy materials. The oxide film formed on the surface of zirconium alloy members has an n-type semiconductor characteristic of oxygen-deficient type (ZrO.sub.2-x). The oxygen vacancy exists as an anion defect in the oxide film. The anion defect is kept electrically neutral by compensation with two electrons. When Fe ions, Ni ions and Cr ions undergo substitution at the sites of Zr ions in the oxide film, an oxygen vacancy is formed (cation vacancy), but the oxygen vacancy (cation vacancy) is not compensated with two electrons to form a cation defect. The two electrons compensating for the anion defect have a higher energy level and are liable to travel according to a potential gradient and thus determine the electron conductivity of the oxide film. On the other hand, the cation defect serves as a trap site for electrons and thus lowers the electron conductivity of the oxide film. Corrosion (oxidation) of zirconium alloy material in reactor core water depends on a balance between the charge transfer by oxygen ions in the film through oxygen vacancies toward the metal side and the charge transfer by electrons from the metal side toward the surface of the oxide film, and thus the corrosion rate is determined by the slower one of the above-mentioned two charge transfers in opposite directions to each other. In the BWR circumstance, the charge transfer by electrons from the metal side toward the surface of the oxide film is a predominant rate-determining factor. When the electron conductivity is lowered by the presence of Fe ions, Ni ions and Cr ions in the oxide film, the corrosion resistance will be increased. In order to substitute Zr ion-occupied positions with Fe ions, Ni ions and Cr ions in the oxide film, it is necessary that Fe, Ni and Cr exist in a solid solution state or exist in fine intermetallic compound phases in the Zr alloy, and larger amounts of these metal elements than those of the currently available zircaloy-2 material are uniformly distributed therein. Hydrogen pickup is due to reaction of Zr with water and absorption of a portion of hydrogen generated by the corrosion reaction into the alloy material. The higher the corrosion resistance, the smaller the amount of the generated hydrogen and the lower the hydrogen pickup ratio. Nodular corrosion is a phenomenon that the above-mentioned corrosion reaction locally proceeds due to the local deficiency of substituent Fe ions, Ni ions and Cr ions at the Zr-occupied positions in the oxide film. In order to prevent such a deficiency, it is necessary to uniformly distribute these elements throughout the alloy. For uniform distribution of these alloy elements, it is effective to provide a heat treatment to heat the alloy material to a .beta.-phase temperature region and/or (.alpha.+.beta.) phase temperature region and successive quenching as a step in the process. By the heat treatment the intermetallic compound phases containing the alloy elements such as Zr(Fe, Cr).sub.2, Zr(Fe, Ni).sub.2, Zr.sub.2 (Ni, Fe), etc. can be made to have an average grain size of 0.4 .mu.m or less, and the Sn.multidot.Ni intermetallic compound phase can be made to have an average grain size of 0.2 .mu.m or less, and they can be uniformly distributed throughout the alloy. It is effective to make a Fe/Ni ratio 1.4 to 15, preferably 10 or less and not to remove Cr from the alloy. Among the intermetallic compound phases, Zr(Fe, Cr).sub.2 grains (hexagonal crystalline system) are finest, followed by Zr(Fe, Ni).sub.2 grains (cubic crystalline system) and Zr.sub.2 (Ni, Fe) grains (hexagonal crystalline system) are coarsest. By addition of Cr, the number of finest Zr.sub.2 (Fe, Cr).sub.2 grains (hexagonal crystalline system) is increased, and by increasing the Fe/Ni ratio, a ratio of finest (Fe, Ni).sub.2 grains to coarsest Zr.sub.2 (Ni, Fe) grains is increased. Effect of making the intermetallic compound phases finer, and of a uniform distribution, on an increase in the corrosion resistance, will be explained below. When the zirconium alloy material is irradiated with neutrons in a nuclear reactor, the stability of intermetallic compound phases is lowered, and Fe, Ni and Cr are dissolved to form a solid solution in the matrix. As explained above, as a result of forming a solid solution of Fe, Ni and Cr, Fe, Ni and Cr undergo substitution at lattice positions of Zr in the oxide film to lower the electron conductivity. By making the intermetallic compound phases finer, their surface area is increased to promote their dissolution and increase the concentration of Fe, Ni and Cr solid solution. By uniform distribution, the concentration of the solid solution is made uniform to elevate the uniformity of electron conductivity of the oxide film and prevent the nodular corrosion. On these grounds the corrosion resistance (nodular corrosion resistance) and hydrogen pickup resistance of zirconium alloy material can be increased. As a result, it is possible to make the zirconium alloy member thinner. Below 1% by weight of Sn neither sufficient corrosion resistance nor sufficient strength can be obtained, whereas above 2% by weight of Sn, no more remarkable effect can be obtained, but the workability is lowered. Thus, 1 to 2% by weight, particularly 1.2 to 1.7% by weight, of Sn is preferable. At least 0.20% by weight of Fe is required for increasing the hydrogen pickup resistance. Above 0.35% by weight of Fe, no more remarkable effect can be obtained, but the workability is lowered. Thus, not more than 0.35% by weight, particularly 0.22 to 0.30% by weight of Fe is preferable. A very small amount of Ni, i.e. at least 0.03% by weight of Ni, is contained in order to remarkably increase the corrosion resistance, but Ni promotes hydrogen pickup, resulting in an increase in the embrittlement. Thus, not more than 0.16%, particularly 0.05 to 0.10%, by weight of Ni is preferable. The present Zr-based alloy can contain 0.05 to 0.15% by weight of Cr. At least 0.05% by weight of Cr is required for increasing the corrosion resistance and strength, whereas, above 0.15% by weight of Cr, the workability is lowered. Thus, 0.05 to 0.15% by weight of Cr is preferable. The present Zr-based alloy can be used for cladding tubes, spacers, channel boxes, and water rods. With the present Zr-based alloy, the former three members can have an average burnup level of 50 to 550 GWd/t. Even in that case, zircaloy-2 alloy can be used for the water rods. The Zr-based alloy for use in the present fuel assembly further includes zircaloy-2 (Ti: 1.2-1.7 wt. %; Fe: 0.07-0.20 wt. %; Cr: 0.05-0.15 wt. %, Ni: 0.03-0.08 wt. %, the balance: substantially Zr), and zircaloy-4 (Ti: 1.2-1.7 wt. %; Fe: 0.18-0.24 wt. %; Ni: <0.07 wt. %); the balance: substantially Zr), and these alloys can be used in combination of the afore-mentioned alloy in view of the average discharge burnup level. Cladding tubes for use in the present invention are preferably those prepared by quenching from (.alpha.+.beta.) phase region or .beta.-phase temperature region after the final hot plastic working and successive repetitions of cold plastic working and annealing. Particularly, quenching from the (.alpha.+.beta.) phase temperature region is preferable, because the successive cold plastic working is better than that quenched from the .beta.-phase temperature region. Preferable Zr-based alloys are those quenched from .beta.-phase or (.alpha.+.beta.) phase temperature region, and the quenching treatment is preferably carried out after the final hot plastic working, but before the final cold plastic working, and particularly preferably before the initial cold plastic working. Preferable (.alpha.+.beta.) phase temperature region is in a range of 800.degree. to 950.degree. C., and preferable .beta.-phase temperature region is in a range of 950.degree. to 1,100.degree. C. Quenching is carried out from these temperature regions with flowing water, sprayed water, etc. Particularly preferably, quenching is carried out before the initial cold plastic working, where it is preferable to conduct local heating by high frequency heating to the outer periphery while passing water into the tube shell. As a result, the inner surface of the tube is not hardened and the ductility is increased, whereas the outer surface of the tube is hardened and the corrosion resistance is increased and the hydrogen pickup ratio is lowered. Since heating in the (.alpha.+.beta.) phase temperature region produces different proportions of the .alpha.-phase to the .beta.-phase, depending on a temperature, it is preferable to select a temperature at which the .beta.-phase is mainly formed. The .alpha.-phase is not converted by the quenching, contributing to a lower hardness and a higher ductility, and quenching from the region converted to the .beta.-phase forms a needle-like phase of high hardness with a low cold workability. However, existence of even a small proportion of the .alpha.-phase can give a high cold plastic workability, and low corrosion resistance and hydrogen pickup ratio. It is preferable to conduct heating at a temperature, where the .beta.-phase has an area ratio of 80 to 95%, and quenching from that temperature. Heating is carried out for a short time, for example, for not more than 5 minutes, particularly 5 seconds to one minute. Prolonged heating is not preferable because it allows crystal grains to grow, forming precipitates and lowering the corrosion resistance. Annealing temperature after the cold plastic working is preferably 500.degree. to 700.degree. C., particularly preferably 550.degree. to 640.degree. C. Below 640.degree. C., a higher corrosion resistance can be obtained. It is preferable to conduct the heating in an Ar atmosphere or in high vacuum. The vacuum degree is preferably 10.sup.-4 to 10.sup.-5 Torr, and it is preferable that no substantial oxide film is formed on the alloy surface by annealing and the alloy surface shows an uncolored metallic luster. The annealing time is preferably 1 to 5 hours. It is preferable to conduct welding by TIG, laser beam or electron beam, and particularly by TIG. It is also preferable that the end plugs and the cladding tubes are preferably made from a Zr-based alloy material of the same composition, and a He gas is used for sealing at a high pressure depending on a desired burnup level, for example, 3 to 20 atmospheric pressures. (3) Combination of Zr-based alloy materials with treatments for respective burnup levels: (a) Burnup level of 50 to 55 GWd/t: The above-mentioned highly corrosion-resistant, highly Fe--Ni, zirconium-based alloy is used for cladding tubes, spacers, and a channel box, where the cladding tubes and spacers are hardened in the above-mentioned (.alpha.+.beta.) phase temperature region or .beta.-phase temperature region, and the channel box is subjected to a heat treatment in the .beta.-phase temperature region to bring the crystallographic orientations into a random distribution. Zircaloy-2 is used for the water rods and is subjected to a heat treatment in the .beta.-phase temperature region to bring the crystallographic orientations into a random distribution. The water rods have an axial distribution of larger wall thickness and a large thickness at the corners. The water rods are connected and fixed to upper and lower tie plates. (b) Burnup level of 45 GWd/t: Zircaloy-2, zirconium-based alloy is used for cladding tubes, a channel box and water rods, where the cladding tubes are subjected to the above-mentioned (.alpha.+.beta.) or .beta.-phase heat treatment, and the channel box and the water rods are subjected to a heat treatment in the .beta.-phase temperature region to bring the crystallographic orientations in a random distribution. Highly corrosion-resistant, highly Fe--Ni, zirconium-based alloy is used for spacers and is subjected to (.alpha.+.beta.) or .beta.-phase hardening. The channel box having a large wall thickness at the corners is used. (c) Burnup level of 38 GWd/t: Zircaloy-2 alloy is used for cladding tubes, spacers, a channel box and water rods, where the cladding tubes and the spacers are hardened in the (.alpha.+.beta.) or .beta.-phase temperature region, and the channel box is heat treated in the .beta.-phase temperature region to bring the crystallographic orientations into a random distribution. A straight channel box is used. It is preferable to heat-treat the water rods in the .beta.-phase temperature region to bring the crystallographic orientations in a random distribution. (d) Burnup level of 32 GWd/t: Zircaloy-2 is used for cladding tubes, and zircaloy-4 is used for other members. Zircaloy-2 can be used for a channel box and spacers. The cladding tubes and the spacers are hardened in the (.alpha.+.beta.) phase temperature region or the .beta.-phase temperature region, and the channel box is heat-treated in the .beta.-phase temperature region to bring the crystallographic orientations into a random distribution and is a straight one with a uniform wall thickness. The water rods can be likewise heat treated in the .beta.-phase temperature region to bring the crystallographic orientations into a random distribution and the water rods subjected to such a treatment are preferably used. Spacers of a lattice type can be used, where plate-shaped materials are welded into the lattice form. Thus, the hardening is carried out in view of the plate-shaped materials and at least one run of each of cold plastic working and annealing must be carried out after the hardening. In the following, the present invention will be described in terms of examples. These examples are illustrative, and not limiting, of the present invention. PREFERRED EMBODIMENTS OF THE INVENTION Example 1 (1) FIG. 1 is a cross-sectional view of a fuel assembly for a boiling water nuclear reactor according to the present invention. A BWR fuel assembly comprises a large number of fuel rods 1, spacers 7 provided at a plurality of stages for supporting the fuel rods 1, each fuel rod loaded with fuel pellets in a cladding tube, at desired distances from one another, a channel box 4 of square cylinder for encasing the fuel rods and the spacers, an upper tie plate 5 and a lower tie plate 6 for supporting the fuel rods 1 at both ends, respectively, water rods 2 provided at the center region of the spacers, and a handle 11 for carrying the entire assembly, as shown in FIG. 1. The fuel assembly can be fabricated through the ordinary steps. The fuel channel box 4 encases the fuel rods 1 and the water rods 2 assembled by the fuel spacers 7, and the upper tie plate 5 and the lower tie plate 6 are fixed by the water rods 2. The fuel channel box 4 is in a shape of a square cylinder, prepared by joining two open channel-shaped (U-shaped) plate members by plasma welding. The channel box rectifies a stream of steam generated on the surfaces of fuel rods and a stream of high temperature water passing through the clearances among the fuel rods and acts to guide the streams upwards forcedly during the reactor operation. Since the inner pressure is slightly higher than the external pressure, the channel box in use is under a stress expanding the square cylinder outwards for a long time. In the present fuel assembly three water rods 2 are provided symmetrically to one another in the center region of the spacers 7 and are each fixed to the tie plates by screw means 3 at both ends. The channel box 4 is fixed to the upper tie plate 5 by screw means and the entire fuel assembly can be carried by the handle 11. In this Example, the fuel rods are not fixed to the tie plates. (2) The channel box is heat-treated so that the crystallographic orientation parameter in the normal-to plate direction of <0002> crystallographic orientation (Fr value) can be 0.25 to 0.5, the crystallographic orientation parameter in the longitudinal direction (Fl value) 0.25 to 0.36 and the crystallographic orientation parameter in the width direction (Ft value) 0.25 to 0.36. By making such an orientation by the heat treatment, .beta.-Zr crystal grain size will be 50 to 300 .mu.m on average and the irradiation elongation can be remarkably prevented, whereby an interference between the channel box and the control rods can be prevented. FIG. 2 is a perspective view showing one embodiment of fabricating a channel box according to the present invention. Two zircaloy-C plates having an alloy composition shown in Table 1 were cold bent to open channel-shaped plates to obtain two open channel-shaped members having a length of 4 m, and the open channel-shaped member were butt-welded to each other along the channel edges by laser or plasma welding to form a square cylinder 12. Projections on the welding seams 17 were made flat by finishing. Then, the square cylinder 12 was heated to a .beta.-phase temperature region by high frequency induction heating and successively quenched with cooling water injected from nozzles 16 provided below a high frequency induction heating coil 14. The square cylinder 12 was passed through the coil 14 at a constant speed from the upside downwards, whereby the entire heat treatment was completed. Feeding speed of the square cylinder 12 and the power output of a high frequency power source 15 were so adjusted that the heating temperature could be 1,100.degree. C. and the retention time at 980.degree. C. or higher could be at least 10 seconds. After the heat treatment, test pieces, 40 mm wide and 40 mm long, were cut out from the square cylinder to measure F values. Table 2 shows the results of the measurement. Heat treatment parameter (P) was 1.96 and the heat treatment was carried out by fixing an austenite stainless steel mandrel 18 to the square cylinder 12 at both ends by screw means 13. As is apparent from Table 2, the <0002> bottom face and <1010> column face of the hexagonal column had Fr, Fl, and Ft values each of substantially 1/3 as F values and were in a completely random crystallographic orientation. The square cylinder had a .beta.-Zr crystal grain size of about 100 .mu.m on the average. After the heat treatment, the square cylinder was reshaped with a high dimensional precision and subjected to sand blasting and pickling to remove the surface oxide film, and then subjected to an autoclave treatment with steam. TABLE 1 ______________________________________ Alloy Alloy element species Sn Fe Cr Ni O Zr ______________________________________ Zircaloy-4 1.50 0.21 0.10 -- 0.12 bal. Zircaloy-2 1.50 0.15 0.10 0.10 0.12 bal. Zircaloy-C 1.50 0.25 0.10 0.10 0.12 bal. ______________________________________ TABLE 2 ______________________________________ Heat (0002) face (1010) face treatment Fr Fl Ft Fr Fl Ft ______________________________________ 1100.degree. C./10s 0.333 0.333 0.334 0.333 0.334 0.333 ______________________________________ FIGS. 3A, 4A and 5 are perspective views showing channel boxes having different cross-sectional profiles in the longitudinal direction, respectively, and FIGS. 3B and 4B are cross-sectional views of FIGS. 3A and 4A each at an intermediate level. FIG. 4C shows a modification of the profile shown in FIG. 4B. The side surfaces and corner edges of the channel box shown in FIG. 3A have a uniform wall thickness distribution throughout the longitudinal direction, as shown in FIG. 3B, whereas the side surfaces of the channel box shown in FIG. 4A have recesses 21 having a smaller wall thickness on the outer surfaces than that of the corner edges 20, as shown in FIG. 4B. The recesses can be formed on the inner surface sides of the channel box as shown in FIG. 4C. The side surfaces of the channel box shown in FIG. 5 have a staged wall thickness distribution in the longitudinal direction, that is, there are recesses of different wall thicknesses on the outer surface sides, i.e. smaller wall thickness at the upper part 22 than at the lower part 23 as shown in FIG. 5. Furthermore, the corner edges 20 of the channel box as shown in FIG. 5 have a largest wall thickness among the wall thicknesses of the side surfaces. The recesses on the side surfaces can be formed by chemical etching using an aqueous acid solution of hydrogen fluoride and nitric acid or by mechanical working. In this Example, the outer side surface were mechanically worked to form recesses thereon. (3) FIG. 6 is a partially cutaway view of a fuel rod according to the present invention. The present fuel rod comprises a cladding tube 24, fuel pellets 25 loaded in the cladding tube 24, and end plugs 27 and a prenum spring 26, and a helium (He) gas being filled inside. In this Example, He is sealed therein at 15-25 atmospheric pressures. The cladding tube 24 is fabricated in the following manner: A pure Zr liner is provided on the inside surface of the cladding tube 24. The liner is provided onto the inner surface of a tube shell after the heat treatment and then subjected to cold plastic working and annealing. As the tube shells, tube shells having an outer diameter of 63.5 mm and a thickness of 10.9 mm, made from the alloys shown in Table 3 by hot rolling were used. Each tube shell was passed through a high frequency induction heating coil and heated, while passing water into the tube shell from the bottom side upwards, and then quenched by injecting water onto the outer surface of the tube shell from nozzles provided just below the coil. The maximum heating temperature was 930.degree. C., which falls in the (.alpha.+.beta.) phase temperature region, and an average cooling speed was about 150.degree. C./s from 930.degree. C. to 500.degree. C. The high frequency-hardened tube shells were subjected to three runs of cold rolling by a Pilger mill and successive annealing at 600.degree. C. in vacuum, where the final annealing temperature was 577.degree. C. Then, the tube shells were used as materials for the fuel cladding tubes and round cells for spacers. In case of the round cells for spacers the tube shells were heated without passing water into the tube shells. Differences in the shapes between the fuel cladding tube and the round cells for spacers are in the tube diameter and the wall thickness, and thus two kinds of tubes having different tube diameters and wall thicknesses were prepared by changing the rolling degree in the final cold rolling. The outer diameter of the round cells for spacers was larger than that of the cladding tube and the wall thickness of the former was smaller than that of the latter. Percent cross-sectional area reduction in the cold rolling was set to 70-80% per run of rolling. The thickness of the liner was about 10-100 .mu.m. The inner surface of the thus obtained cladding tubes had a specific crystallographic orientation, i.e., a Fr value of 0.6 to 0.7. Slip-shaped test pieces were cut out from the thus prepared two kinds of tube shells, and exposed to a higher temperature and pressure steam (500.degree. C.; 10.3 MPa) for 24 hours to investigate weight increases due to corrosion and appearance of corroded test pieces. The results are shown in Table 4. TABLE 3 ______________________________________ Alloy element Alloy species Sn Fe Cr Ni O Zr Fe/Ni ratio ______________________________________ Zircaloy-2 1.50 0.15 0.10 0.05 0.11 bal. 3.0 Zircaloy-A 1.50 0.23 0.10 0.05 0.11 bal. 4.6 Zircaloy-B 1.50 0.23 0.10 0.09 0.11 bal. 2.6 Zircaloy-C 1.50 0.13 0.10 0.09 0.11 bal. 1.4 Zircaloy-D 1.50 1.10 -- 0.08 0.11 bal. 1.3 ______________________________________ TABLE 4 ______________________________________ Increase in Alloy species Appearance corrosion (mg/dm.sup.2) ______________________________________ Zircaloy-2 Partial nodular .about.150 corrosion Zircaloy-A Uniform black 60.about.80 Zircaloy-B Uniform black 60.about.80 Zircaloy-C Uniform black 60.about.80 Zircaloy-D Full nodular .about.250 corrosion ______________________________________ In case of so far used zircaloy-2 and zircaloy-D, Table 3 shows occurrence of nodular corrosion and a high increase in corrosion. In case pf alloys (zircaloy-A to zircaloy-C) having a higher Fe/Ni ratio than 1.4, and higher Fe and Ni contents than those set in the Standard Code range for zircaloy-2, no nodular corrosion appeared, but black oxide films of uniform thickness were formed, showing a very high corrosion resistance. It can be seen from the foregoing results that alloys having a higher Fe/N ratio than 1.4, containing Cr and having higher Fe and Ni contents than those set in the Standard Code range for zircaloy-2, have a higher corrosion resistance, even if used in a nuclear reactor. Corrosion degree of the alloys after the 6 years' service can be estimated to 130 mg/dm.sup.2 (oxide film thickness: 8 .mu.m) and the spacer can be estimated to have a hydrogen content of less than about 250 ppm. It is preferable for the zirconium-based alloy for the cladding tubes and spacers to precipitate grains of tin-nickel intermetallic compound having grain sizes of not more than 0.2 .mu.m and grains of iron-nickel-zirconium intermetallic compounds having grain sizes of 0.1 to 0.5 .mu.m in the .alpha.-phase zirconium crystal grains. In this Example, the grain sizes of the former were as fine as about 0.01 .mu.m. (4) FIGS. 7 and 8 are partially cutaway views of water rods, and in this Example a water rod having a larger diameter as shown in FIG. 8 is used and zircaloy-2 shown in Table 1 is used as the alloy for the tube shell. As explained above, a tube shell is hardened in the (.alpha.+.beta.) phase temperature region or the .beta.-phase temperature region, and then subjected to cold plastic working to a desired shape and successive annealing, whereby a water rod can have smaller diameter parts 28, a larger diameter part 29 and end plug parts 30. Screw means are provided at the end plug parts 30 to fix the end plug parts to the upper and lower tie plates, respectively, as already explained. (5) FIGS. 9 and 10 show plan views of spacers, respectively, and in this Example, round cell-type spacer shown in FIG. 10 is used and zircaloy-B shown in Table 3 is used for the round cell spacer material. FIG. 11 is a perspective view of round cells for spacers, which are prepared by hardening heat treatment of tube shells from the (.alpha.+.beta.) phase temperature region, followed by repetitions of cold plastic working and annealing to obtain tube shells of desired smaller wall thickness, and cutting the thus obtained tube shell, thereby obtaining round cells in the desired shape. (6) Fuel pellets having a uranium-235 enrichment of about 4.5 wt. % are loaded into the above-mentioned fuel rods to obtain a discharge burnup level of 50 to 550 GWd/t. The service duration of the fuel is 6 to 6.5 years. The end plugs are butt-welded to ends of each cladding tube by laser welding. For the tie plates, an austenite steel casting containing not more than 0.03% by weight of C, not more than 2% by weight of Si, not more than 2% by weight of Mn, 8 to 12% by weight of Ni, and 17 to 21% by weight of Cr, the balance being Fe, is used. The casting is subjected to a solubilization treatment at 1,100.degree. C. The above-mentioned channel boxes were subjected to a fast neutron irradiation test and it was found that a strain occurrence was as very small as 0.3.times.10.sup.-4 at 3.times.10.sup.22 n/cm.sup.2. Test Example 1 The alloys shown in Table 1 were subjected to heat treatments shown in Table 5 to investigate relations between the degree of randomness of crystallographic orientation and elongation by neutron irradiation of Zr-based alloys for use in the above-mentioned channel box. The degree of randomness of the crystallographic orientation was changed by changing the heat treatment conditions. TABLE 5 ______________________________________ Heat Max. Retention Cooling treatment heating temp. time speed No. (.degree.C.) (sec.) (.degree.C./sec.) P ______________________________________ 1 No heat treatment 2 900 (.alpha. + .beta.) 600 200 -- 3 1000 (.beta.) 60 200 2.24 4 1000 (.beta.) 600 200 3.54 5 1000 (.beta.) 60 150 4.03 6 1000 (.beta.) 5 200 0.84 ______________________________________ All of these alloys were in a plate shape having a thickness of 2 mm, and cold rolling and annealing at 650.degree. C. for 2 hours were repeatedly carried out for the alloys before the heat treatment. Heat treatment Nos. 2 to 5 shown in Table 5 were carried out by cutting out test pieces, 40 mm wide and 40 mm long, from the test plates, heating the test pieces in an electric furnace, followed by cooling in water. Heat treatment No. 6 was carried out by retaining a test piece cut out from the test plate in an infrared heating furnace, followed by cooling in water. Parameter P was calculated according to the above-mentioned equation. Table 6 shows a result of measuring F values of (0002) bottom face [the face in parallel with the (0001) face] and (1010) column face [the face normal to the (0001) face] of hexagonal columns of test pieces subjected to heat treatment Nos. 1 to 6. Fr values show an orientation probability in the normal-to-plate direction, Fl values that in the plate-rolling direction, and Ft value that in the direction perpendicular to the former two. No difference was observed in the F values due to changes in the alloy compositions. It can be seen from Table 6 that the test piece (Heat treatment No. 1) prepared by repetitions of ordinary cold rolling and annealing had a high Fr value such as about 0.7 on the (0002) face and a low Fr value such as about 0.15 on the (1010) column face, and thus the (0002) face was oriented substantially in parallel with the plate surface. The test piece (Heat treatment No. 2) subjected to heating to the (.alpha.+.beta.) phase temperature region of 900.degree. C., followed by cooling in water, had F values substantially equal to those of the test piece without heat treatment (Heat treatment No. 1), and it can be seen therefrom that the crystallographic orientation was not substantially changed by the heating to the (.alpha.+.beta.) temperature region, followed by cooling. Test piece (Heat treatment No. 6) subjected to heating up to 1,000.degree. C. and retention in the .beta.-phase temperature region (>980.degree. C.) for 5 seconds had a decrease in the Fr value and an increase in the Fl value and the Ft value on the (0002) face, but an increase in the Fr value and a decrease in the Fl value and the Ft value on the (1010) column face. It can be seen therefrom that the crystallographic orientation was brought into a random distribution and the Fl value was made higher than 0.20 by making the P value higher than 0.8, but failed to satisfy the target value for preventing an interference between the channel box and the control rods when placed in the core periphery region, that is, Fl value on the (0002) face .gtoreq.0.25. Test pieces of Heat treatment Nos. 3 to 5 all satisfied a Fl value .gtoreq.0.25 and it can be seen therefrom that the channel box and the control rods will not interfere with one another when placed in the core peripheral region. TABLE 6 ______________________________________ Heat (0002) face (1010) face treatment No. Fr Fl Ft Fr Fl Ft ______________________________________ 1 0.672 0.108 0.220 0.158 0.448 0.393 2 0.666 0.124 0.210 0.156 0.445 0.398 3 0.414 0.295 0.292 0.301 0.354 0.345 4 0.335 0.352 0.318 0 325 0.329 0.344 5 0.336 0.334 0.330 0.330 0.335 0.335 6 0.470 0.203 0.327 0.209 0.401 0.390 ______________________________________ Strain developed due to neutron irradiation growth was investigated by changing Fr values in the above-mentioned heat treatment. FIG. 12 is a diagram showing relations between the fast neutron irradiation dosage and strain developed due to the irradiation growth, where calculation results of influence of Fr value on the curving degree of channel boxes when exposed to one cycle of irradiation in the core periphery region and 4 cycles of irradiation in the core center region are summarized. As shown in FIG. 12, when the Fr value exceeds 0.4, strain rapidly increases with increasing neutron irradiation dosage, but below 0.4, the strain is saturated with no more increase even if irradiated. Particularly at Fr=0.35, the <0001> crystallographic orientation is substantially in a random distribution, and thus the strains in the normal-to-plate direction, the longitudinal direction and the width direction are off-set among the individual crystals and thus the strain developed is less than 0.5.times.10.sup.-4, that is, no strain develops at all. At Fr=0.4, the strain developed is small up to an irradiation dosage of 3.times.10.sup.22 n/cm.sup.2, with gradual increase with a higher neutron irradiation dosage. At Fr=0.35, the strain never increases with increasing neutron irradiation dosage. Relations between Fr values and strains due to irradiation growth by irradiation of fast neutrons at a dosage of 3.times.10.sup.22 n/cm.sup.2 were investigated and it was found that the strain was rapidly increased with increasing Fr values, and particularly a strain due to the irradiation growth at Fr=0.35 was about 0.2.times.10.sup.-4, which was considerably smaller by about 1/7 than about 1.5.times.10.sup.-4 at Fr=0.4, which was considerably lower by about 1/3 than that at Fr=0.5. The strain at Fr=0.5 was about one-half of that at Fr=0.6, which was about one-half of that at Fr=0.7. No remarkable effect was obtained above Fr=0.4. Round crystal grains observed in the metallic structures of the heat-treated test pieces Nos. 1, 3 and 4 contained no .alpha.-Zr crystal grains. Observed polygonal crystal grains were .beta.-Zr crystal grains formed by heating to and retaining in the .beta.-phase temperature region, and with increasing retention time at 1,000.degree. C. from one minute to 10 minutes the .beta.-Zr crystal grains grew to larger grain sizes. Laminar or needle-like structures observed in the .beta.-Zr crystal grains were formed when the .beta.-Zr crystal grains was transformed again to the .alpha.-Zr crystal grains in the cooling step without corresponding to the .beta.-Zr crystal grain boundaries. Relations between the .beta.-Zr crystal grain sizes and the Fr values on the (0002) face reveal that when the .beta.-Zr crystal grain size exceeds 200 .mu.m, an aggregated structure having a Fr value of not more than 0.35 will be formed. By making the crystal grains grow in this manner, the crystallographic orientation of (0002) face can be brought in a random distribution, and the degree of randomness of that orientation is about 75% at a Fr value of 0.40 and a Fl value of 0.30, where the grain size will be about 100 .mu.m. With crystal grain size of 150 .mu.m or more, a degree of randomness will be about 80% or more, and the Fr value will be 0.385 at a Fl value of 0.320, and at a Fr value of 0.35 and a Fl value of 0.34, the degree of randomness will be about 90% or more, where the crystal grain size will be about 250 .mu.m or more. Relations between the .beta.-Zr crystal grain sizes and the strain due to the irradiation growth reveal that with grain sizes of 20 .mu.m or more, the strain will be 4.times.10.sup.-4 or less, and with grain sizes of 90 .mu.m or more, the strain will be considerably as small as about 1.5.times.10.sup.-4. With grain sizes of 150 .mu.m or more, the strain will be very small, for example, 0.5.times.10.sup.-4 or less. Particularly with grain sizes of 200 .mu.m or more, the strain will be about 0.3.times.10.sup.-4. FIG. 13 is a diagram showing relationships for Fr values of alloys shown in Tables 1 and 3, between a temperature and a retention time. As shown in FIG. 13, below 980.degree. C., the Fr value will be not more than 0.20, and the <0002> crystallographic orientation will be hardly brought into a random distribution. By heating in a region formed between a straightly elevating line at 980.degree. C. for at least 11 seconds upward and a straight line connecting a point at 980.degree. C. for 11 seconds to a point at 1,240.degree. C. for 1.1 second, a Fr value of not less than 0.25 can be obtained and a higher degree of randomness can be obtained. Furthermore, by heating in a region formed by a straightly elevating line at 980.degree. C. for at least 6 seconds and a straight line connecting a point at 980.degree. C. for 6 seconds to a point at 1,240.degree. C. for 0.6 seconds, a Fr value of more than 0.20 but less than 0.25 can be obtained. By heating outside the latter region, a Fr value will be less than 0.20 and the degree of randomness is lower, resulting in less effect on the expansion. Relations between the heating temperature and the time can be represented by the following parameter, which seems to be valid up to about 1,200.degree. C. From the relationships between parameter P=(3.5+logt).times.log(T-980) and strain due to the irradiation growth, it can be seen that the strain due to the irradiation growth largely depends on the parameter P determined by relations between the temperature and the retention time in the heat treatment. Parameter P is an important factor for determining the crystallographic orientation parameter in the Zr <0001> crystallographic orientation. When the P value is more than 0.5, the strain due to the irradiation growth is suddenly decreased, and when the P value is between 0.5 and 3.5 the strain is gradually decreased. When the P value is more than 3.5, the strain will be substantially constant and nearly zero. Particularly when the P value is 1.5 or more, a remarkable effect can be obtained, and 3.0 to 5 is preferable for the P value. Test Example 2 Sponge zirconium was melted in vacuum to form alloy ingots of various compositions containing about 1.5% by weight of Sn, 0.10 to 0.50% by weight of Fe, 0 to 0.30% by weight of Ni, and 0.08 to 0.13% by weight of Cr, the balance being substantially Zr. The ingots were hot rolled (700.degree. C.) and annealed (700.degree. C. for 4 hours), retained in the (.alpha.+.beta.) phase temperature region (900.degree. C.) or the .beta.-phase temperature region (1,000.degree. C.) for 2 to 3 minutes and then cooled with water. Then, the ingots were subjected to three repetitions of cold rolling (rolling degree for one run: 40%) and successive intermediate annealing at 600.degree. C. for 2 hours to make plates having a thickness of 1 mm. Corrosion tests were carried out by retaining test pieces in high temperature steam at a pressure of 10.3 MPa and 410.degree. C. for 8 hours and then at 510.degree. C. and the same pressure as above for 16 hours, and measuring a corrosion increment. To investigate influences of material compositions on corrosion in an accelerated manner, the test pieces were heated at 530.degree. C., 620.degree. C. and 730.degree. C. each for two hours. As to the hydrogen pickup characteristics, such principles can be employed that reaction of Zr with water produces an oxide (ZrO.sub.2) and also generates a hydrogen gas at the same time, and moles of water that have reacted with zircaloy can be obtained and also moles of hydrogen gas generated correspondingly can be obtained by measuring a weight increase of test pieces due to the oxidation. Thus, a hydrogen pickup ratio can be determined by measuring the amount of hydrogen in each test piece after the corrosion test by chemical analysis, calculating moles of absorbed hydrogen to obtain a ratio of moles of absorbed hydrogen to that of the generated hydrogen. FIG. 14 shows the generation or no generation of nodular corrosion, wherein the round mark ".largecircle." shows no generation of nodular corrosion on the surface or side surfaces, irrespective of final annealing temperature, where the corrosion increment was less than 45 mg/dm.sup.2, and the crossed mark ".times." shows generation of nodular corrosion on the surface and the side surfaces, where the corrosion increment exceeded 50 mg/dm.sup.2. Numerical FIGURES in FIG. 14 show corrosion increments. It can be seen from FIG. 14 that alloy compositions that can prevent generation of nodular corrosion exist in a region on higher Ni and Fe content sides, partitioned by the dotted line in FIG. 14. The dotted line was obtained from such an equation as 0.15.times.Fe content (wt. %) +0.25.times.Ni content (wt. %)=0.375. FIG. 15 is a diagram showing influences of Fe and Ni contents on a corrosion increment. As shown in FIG. 15, corrosion in high temperature and high pressure water could be considerably reduced by increasing Fe and Ni contents. Particularly, the corrosion increment could be rapidly reduced by addition of a very small amount of Ni. By addition of at least 0.06% by weight of Ni at about 0.10% by weight of Fe, at least 0.04% by weight of Ni at about 0.15% by weight of Fe and 0.03% by weight of Ni at 0.21% by weight of Fe, the corrosion increment could be made less than 45 mg/dm.sup.2, and no nodular corrosion was generated. FIG. 16 shows influences of Fe content on the hydrogen pickup ratio, where the triangular mark ".DELTA." shows a hydrogen pickup ratio of alloys containing 0.11% by weight of Ni, and the round mark ".largecircle." shows that of alloys containing 0.05% by weight of Ni. In FIG. 16, the dotted lines shows test results of alloys without quenching from the (.alpha.+.beta.) phase temperature region or from the .beta.-phase temperature region. Full lines show hydrogen pickup ratios of the alloys with quenching from the (.alpha.+.beta.) temperature region in the heat treatment. From FIG. 16 it can be seen that the hydrogen pickup ratio can be made less than 11% by quenching from the (.alpha.+.beta.) temperature region, and the hydrogen pickup ratio can be reduced by making the Fe content 0.21% by weight or higher, irrespective of the Ni content. FIG. 17 shows influences of Fe and Ni contents on the hydrogen pickup ratio. Below a Ni content of less than 0.16% by weight, the hydrogen pickup ratio is as low as 11%, whereas above 0.2% by weight of Ni, the hydrogen pickup ratio is suddenly increased to reach 40%. Thus, the Ni content must be 0.15% or less by weight. By making the Fe content 0.21% by weight or more, a hydrogen pickup ratio of not more than 10% can be obtained. FIG. 18 is a diagram showing influences of (Fe/Ni) ratios on the hydrogen pickup ratio. As shown in FIG. 18, the round mark ".largecircle." each show a Fe content of less than 0.21% by weight and also show no influences of (Fe/Ni) ratios. At Fe contents of 0.20% by weight or higher, the (Fe/Ni) ratio must be 1.4 or more. As explained above, Fe and Ni have quite reversed actions on the effect on the hydrogen pickup ratio, and thus a ratio of these elements is important. Below a Fe content of 0.2% by weight and above a Ni content of 0.2% by weight, there is no correlation between these elements, but when these contents are reversed to each other, a correlation is observable therebetween. Alloys having increased Fe contents up to 0.48% by weight have a corrosion increment of 43 mg/dm.sup.2 and a hydrogen pickup ratio of 12%. From the viewpoints of corrosion resistance and hydrogen pickup, the Fe content must be increased from 0.21% by weight to about 0.5% by weight, so long as the Ni content is less than 0.16% by weight. However, as will be explained later, when the sum total of Ni and Fe contents is as large as 0.64% by weight, the cold plastic workability is abruptly lowered, and thus this is not preferable for the materials that can be made into members having a small thickness by the cold plastic working as explained before. Thus, the sum total of Fe and Ni contents must be less than 0.40% by weight. Inspection of precipitates in the alloys containing 0.25% by weight of Fe and 0.11% by weight of Ni, quenched from the (.alpha.+.beta.) phase temperature region, revealed that there was intermetallic compounds of tin and nickel, which are uniformly distributed and precipitated in the .alpha.-phase zirconium crystal grains. The precipitates were composed of Sn.sub.2 Ni.sub.3 and had very small grain sizes of about 10 nm. No such precipitates were inspected in the same alloys without quenching from the (.alpha.+.beta.) phase temperature region. No precipitates of Sn--Ni intermetallic compounds were found in the quenched alloys from the (.alpha.+.beta.) phase temperature region, which were subjected to hot plastic working after the quenching. Example 2 A fuel assembly as shown in FIG. 1 was fabricated. Differences from Example 1 will be given below. For cladding tubes, zircaloy-2 (Sn: 1-2 wt. %; Fe: 0.05-0.20 wt. %; Cr: 0.05-0.15 wt. %; Ni: 0.03-0.1 wt. %; the balance: Zr) was used. Hot drawn tube shells subjected to final hot plastic working were hardened by passing water into the tube shells in the same manner as in Example 1, whereby a higher solid solution ratio was obtained on the outer surface side than on the inner surface side and a higher corrosion resistance was obtained on the outer surface side. In this Example, a liner of pure Zr was formed on the inner surface side in the same manner as in Example 1. Fuel rods were the same as shown in FIG. 6, where the end plugs 27 were made from the same material as used for the cladding tubes, and after the fuel rods were loaded with nuclear fuel pellets 25 and the end plugs were butt-welded to the fuel rods by laser welding, and the helium gas was sealed therein. Sealing helium gas pressure was about 10 atmospheric pressures and the pellets 25 had an average uranium-235 enrichment of about 4.0% by weight. Hardening treatment of tube shells for the cladding tubes could be carried out in the (.alpha.+B) phase temperature region at any stage, that is, the tube shell stage over to the stage just before the final cold plastic working. Hardening treatment in the .beta.-phase temperature region could be carried out in the same manner as that in the (.alpha.+.beta.) phase temperature region. In any of the (.alpha.+.beta.) phase and .beta.-phase temperature regions it was preferable that the hardening be carried out at the tube shell stage, and in the .beta.-phase temperature region it was preferable to carry out the hardening before the cold plastic working in advance to the final cold plastic working. The annealing temperature after the cold plastic working was preferably 640.degree. to 500.degree. C. For the channel box, an alloy of the same composition as that for the cladding tubes was used, and after heating at 1,100.degree. C. in the .beta.-phase temperature region for 10 seconds in the same manner as above, a water spraying treatment was carried out, whereby the same Fr value, Fl value and Ft value shown in Table 2 were obtained. From a neutron irradiation dosage of 3.times.10.sup.22 n/cm.sup.2, a very small strain such as 0.3.times.10.sup.-4 resulted. In this Example, the channel box wa of such a type as shown in FIG. 4, where the wall thickness was larger at the corners than on the surface sides. FIG. 4B shows a recess on the outer surface side and FIG. 4C shows a recess on the inner surface side, formed by mechanical working or chemical etching. Spacers having the same structure and made from the same material as in Example 1 were used. The material was subjected to hardening in the (.alpha.+.beta.) phase temperature region or the .beta.-phase region in the same manner as in Example 1. Water rods had the same structure and were made from the same material as in Example 1. With the foregoing structure of the fuel assembly, an average discharge burnup level of 45 GWd/t could be obtained, no nodular corrosion was generated on the cladding tubes, spacers, and water rods, and the channel box had a very small expansion. Example 3 A fuel assembly as shown in FIG. 1 was fabricated. Differences from Example 2 are given below. A channel box having a straight structure as shown in FIG. 3 was fabricated from zircaloy-2 in the same manner as in Example 1 and particularly heat treatment was carried out by heating at 1,100.degree. C. for 10 seconds, followed by cooling with water, whereby substantially the same F values as shown in Table 2 were obtained. An average crystal grain size was about 100 .mu.m, and the same treatment after the heat treatment as in Example 1 was carried out. For cladding tubes, spacers and water rods, the same zircaloy-2 as above was used, and a hardening treatment in the (.alpha.+.beta.) phase temperature region or the .beta.-phase temperature region was carried out after the final hot plastic working. End plugs were butt-welded to the ends of each cladding tube and each water rod by TIG welding. The cladding tubes were loaded with nuclear fuel pellets in the same manner as in Example 2, and an average enrichment of uranium-235 for the nuclear fuel was about 3.4% by weight, and the He sealing pressure was about 5 atmospheric pressures. From the thus fabricated fuel assembly, an average discharge burnup level of 38 GWd/t resulted, and the service duration was about 4.5 years. Example 4 A fuel assembly as shown in FIG. 1 was fabricated. Differences from Example 3 are given belows. For cladding tubes and channel box, zircaloy-2 was used, and for spacers and water rods zircaloy-4 (Sn: 1-2 wt. %; Fe: 0.18-0.24 wt. %; Ni: less than 0.01 wt. %; the balance: substantially Zr) was used. The cladding tubes and spacers were hardened in the .beta.-phase or (.alpha.+.beta.) phase temperature region and had a high corrosion resistance. The spacers were of a lattice type as shown in FIG. 9, and were fabricated from hardened plate members by TIG welding. The nuclear fuel pellets had an average uranium-235 enrichment of about 3% and the average discharge burnup level was 33 GWd/t and the service duration was about 4 years. Since the crystallographic orientations of a channel box as a fuel assembly casing were brought in a random distribution, the channel box is less deformed and can be used in a reactor core for a longer service duration. Corrosion resistances of the individual members of the fuel assembly are increased and the hydrogen pickup ratio is considerably reduced. Thus, a higher burnup level of the fuel assembly can be obtained, contributing to reduction in the amount of spent fuels and to an increase in the reliability of the individual members of the fuel assembly. While we have shown and described several embodiments in accordance with the present invention, it is understood that the same is not limited thereto, but is susceptible of numerous changes and modifications as known to those skilled in the art. Therefore, we do not wish to be limited to the details shown and described herein, but intend to cover all such changes and modifications as are encompassed by the scope of the appended claims. |
description | This application is a continuation of U.S. patent application Ser. No. 15/413,770 filed Jan. 24, 2017, which is a continuation of U.S. patent application Ser. No. 13/577,163, filed Aug. 3, 2012, which is a national stage entry of International Application No. PCT/US2011/023952, filed Feb. 7, 2011, which claims the benefit of U.S. Provisional Patent Application No. 61/416,954, filed Nov. 24, 2010, U.S. Provisional Patent Application No. 61/333,551, filed May 11, 2010, and U.S. Provisional Patent Application No. 61/302,069, filed Feb. 5, 2010, the entireties of which are herein incorporated by reference. The present invention relates generally to nuclear reactor systems, and specifically to nuclear reactor systems that utilize natural circulation of the primary coolant in a single-phase, such as pressurized water reactors (“PWRs”). Over recent years, a substantial amount of interest has grown in developing commercially viable PWRs that utilize the phenomenon of natural circulation (also known as thermosiphon effect) to circulate the primary coolant to both cool the nuclear reactor and to vaporize a secondary coolant into motive vapor. CAREM (Argentina) is a 100 MW(e) PWR reactor design with an integrated self-pressurized primary system through which the primary coolant circulation is achieved by natural circulation. The CAREM design incorporates several passive safety systems. The entire primary system including the core, steam generators, primary coolant and steam dome are contained inside a single pressure vessel. The strong negative temperature coefficient of reactivity enhances the self-controlling features. The reactor is practically self-controlled and need for control rod movement is minimized. In order to keep a strong negative temperature coefficient of reactivity during the whole operational cycle, it is not necessary to utilize soluble boron for burn-up compensation. Reactivity compensation for burn-up is obtained with burnable poisons, i.e. gadolinium oxide dispersed in the uranium di-oxide fuel. Primary coolant enters the core from the lower plenum. After being heated the primary coolant exits the core and flows up through the riser to the upper dome. In the upper part, the primary coolant leaves the riser through lateral windows to the external region, then flows down through modular steam generators, decreasing its enthalpy by giving up heat to the secondary coolant in the steam generator. Finally, the primary coolant exits the internal steam generators and flows down through the down-comer to the lower plenum, closing the circuit. CAREM uses once-through straight tube steam generators. Twelve steam generators are arranged in an annular array inside the pressure vessel above the core. The primary coolant flows through the inside of the tubes, and the secondary coolant flows across the outside of the tubes. A shell and two tube plates form the barrier between primary and secondary coolant flow circuits. AST-500 (Russia) is a 500 MW(th) reactor design intended to generate low temperature heat for district heating and hot water supply to cities. AST-500 is a pressurized water reactor with integral layout of the primary components and natural circulation of the primary coolant. Features of the AST-500 reactor include natural circulation of the primary coolant under reduced working parameters and specific features of the integral reactor, such as a built-in steam-gas pressurizer, in-reactor heat exchangers for emergency heat removal, and an external guard vessel. V-500 SKDI *(Russia) is a 500 MW(e) light water integral reactor design with natural circulation of the primary coolant in a vessel with a diameter less than 5 m. The reactor core and the steam generators are contained within the steel pressure vessel (i.e., the reactor pressure vessel). The core has 121 shroudless fuel assemblies having 18 control rod clusters. Thirty six fuel assemblies have burnable poison rods. The hot primary coolant moves from the core through the riser and upper shroud windows into the steam generators located in the downcomer. The coolant flows due to the difference in coolant densities in the downcomer and riser. The pressurizer is connected by two pipelines, to the reactor pressure vessel and the water clean up system. The NHR-200 (China) is a design for providing heat for district heating, industrial processes and seawater desalination. The reactor power is 200 MW(th). The reactor core is located at the bottom of the reactor pressure vessel (RPV). The system pressure is maintained by N2 and steam. The reactor vessel is cylindrical. The RPV is 4.8 m in diameter, 14 m in height, and 197 tons in weight. The guard vessel consists of a cylindrical portion with a diameter of 5 m and an upper cone portion with maximum 7 m in diameter. The guard vessel is 15.1 m in height and 233 tons in weight. The core is cooled by natural circulation in the range from full power operation to residual heat removal. There is a long riser on the core outlet to enhance the natural circulation capacity. The height of the riser is about 6 m. Even in case of interruption of natural circulation in the primary circuit due to a LOCA the residual heat of the core can be transmitted by steam condensed at the uncovered tube surface of the primary heat exchanger. While the aforementioned PWRs utilize natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant, all of these natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service but also subjects the equipment to corrosive conditions. Furthermore, locating the heat exchange equipment within the reactor pressure vessel results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. However, prior to the present invention, the location of the heat exchange equipment within the reactor pressure vessel was likely deemed necessary to achieve the natural circulation of the primary coolant in the PWR cycle. A drawback of other PWRs that exist in the art is the fact that the reactor pressure vessels have penetrations at both the top portion of the reactor pressure vessel and at the bottom portion of the reactor pressure vessel. Still another drawback of existing PWRs is the fact that a substantial length of piping and a large number of joints are used carry the primary coolant from the reactor pressure vessel to the heat exchange equipment, thereby increasing the danger of failure due to a pipe break scenario. These, and other drawbacks, are remedied by the present invention. A nuclear reactor system is presented herein that, in one embodiment, utilizes natural circulation (i.e., thermosiphon) to circulate a primary coolant in a single-phase through a reactor core and a heat exchange sub-system, wherein the heat exchange sub-system is located outside of the nuclear reactor pressure vessel. In some embodiments, the heat exchange sub-system is designed so as to not cause any substantial pressure drop in the flow of the primary coolant within the heat exchange sub-system that is used to vaporize a secondary coolant. In another embodiment, a nuclear reactor system is disclosed in which the reactor core is located below ground and all penetrations into the reactor pressure vessel are located above ground. In certain embodiment, the inventive nuclear reactor system is a PWR system. In one embodiment, the invention can be a natural circulation nuclear reactor system comprising: a reactor pressure vessel having an internal cavity; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel; a heat exchange sub-system located outside of the reactor pressure vessel; a closed-loop primary coolant circuit that flows a primary coolant through the reactor pressure vessel to cool the reactor core and through the heat exchange sub-system to transfer heat to a secondary coolant; and wherein operation of the reactor core causes natural circulation of the primary coolant through the closed-loop primary coolant circuit in a single phase. In another embodiment, the invention can be a nuclear reactor system comprising: an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis, a major portion of the axial length of the reactor pressure vessel located below a ground level; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor and below the ground level; the reactor pressure vessel comprising a primary coolant outlet port located above the ground level; the reactor pressure vessel comprising a primary coolant inlet port located above the ground level; a heat exchange sub-system located outside of the reactor pressure vessel and above the ground level, an incoming hot leg of the heat exchange system fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the heat exchange system fluidly coupled to the primary coolant inlet port; and wherein the major portion of the reactor pressure vessel is free of penetrations. In yet another embodiment, the invention can be a nuclear reactor system comprising: an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor; a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway; the reactor pressure vessel comprising a primary coolant outlet port in fluid communication with a top portion of the primary coolant riser passageway; the reactor pressure vessel comprising a primary coolant inlet port in fluid communication with a top portion of the primary downcomer riser passageway; at least one steam generator located outside of the reactor pressure vessel, an incoming hot leg of the steam generator fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the steam generator fluidly coupled to the primary coolant inlet port; and wherein the steam generator does not cause any substantial pressure drop in a flow of the primary coolant through the steam generator resulting from an increase in elevation. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. Prior to discussing FIGS. 1-5 in detail, an overview of one specific embodiment of the inventive natural circulation reactor system, and its operation, will be set forth. Those skilled in the art will appreciate that the overview is directed to one very specific embodiment and that the details thereof are not limiting of the present invention in all embodiments. Furthermore, those skilled in the art will appreciate how the overview applies to the subsequent detailed discussion of FIGS. 1-5. I. Overview of One Potential Commercial Embodiment The inventive nuclear reactor system, in one potential commercial embodiment, is a 145 MWe nuclear reactor designed to provide an economical and safe source of clean energy from nuclear fission. Strengths of the inventive nuclear reactor system include its inherent safety and simplicity of operation. The operational simplicity of the inventive nuclear reactor system and the modest outlay required to establish and commission it will make it possible to deliver the fruits of pollution-free nuclear energy to the vast mass of humanity around the globe that does not presently have access to a reliable source of power or to a robust electrical energy delivery system. Competitive with large nuclear reactors on a per-megawatt basis, the inventive nuclear reactor system is tailored to add generation capacity to the installed base incrementally with incremental capital outlays. Due to its inherent operational simplicity, the inventive nuclear reactor system requires a minimal cadre of trained personnel to run the plant. Multiple units of the inventive nuclear reactor system can be clustered at one location or geographically dispersed without a significant increase in the per-megawatt construction cost. Geographical dispersal and underground configuration serve as natural antidotes to post-9/11 concerns. The modest power output of the inventive nuclear reactor system makes it a viable candidate source of reliable electrical energy or for providing heating steam to a city or process steam as a cogeneration plant serving an industrial plant. As a passive small modular reactor of the PWR genre with safety, ease of maintenance and superb security, the inventive nuclear reactor system is ideally suited to serve as a reliable power source to strategic national assets of any country. Design features of the inventive nuclear reactor system that speak to its inherent safety and reliability are: 1. Reactor Core Deep Underground The reactor core resides deep underground in a thick-walled reactor pressure vessel (RPV) made of an ASME Code material that has decades of proven efficacy in maintaining reactor integrity in large PWR and BWR reactors. All surfaces wetted by the reactor coolant are made of stainless steel or Inconel, which eliminates a major source of crud accumulation in the reactor vessel. 2. Natural Circulation of the Reactor Coolant The inventive nuclear reactor system does not rely on any active components, such as a reactor coolant pump, for circulating the primary coolant through the closed-loop primary coolant circuit, which includes flow through the reactor core and the heat exchange sub-system. Instead, the flow of the primary coolant through the reactor pressure vessel, the horizontal steam generators, and other miscellaneous equipment occurs by the pressure head created by density differences in the flowing water in the hot and cold segments of the closed-loop primary coolant circuit. The reliability of gravity as a motive force underpins inherent safety of the inventive nuclear reactor system. The movement of the primary coolant requires no pumps, valves, or moving machinery of any kind, in certain embodiments. 3. No Reliance on Off-Site Power Offsite power is not essential for shutting down the inventive nuclear reactor system. The rejection of reactor residual heat during the shutdown also occurs by natural circulation. Thus, the need for an emergency shutdown power supply at the site—a major concern for nuclear plants—is eliminated. 4. Assurance of a Large Inventory of Water Around and Over the Reactor Core The reactor pressure vessel of the inventive nuclear reactor system has no penetrations in its below-ground portion, which can be the bottom 100 feet, which means that the reactor core will remain submerged in a large inventory of water. All penetrations in the reactor pressure vessel are located in the above-ground portion, or top portion, of the reactor pressure vessel and are small in size. The absence of large piping in the closed-circuit primary coolant circuit precludes the potential of a “large break” LOCA event. 5. All Critical Components Readily Accessible Both the heat exchange sub-system, which includes the steam generators, and the control rod drive system are located outside the reactor pressure vessel at a level that facilitates easy access, making their preventive maintenance and repair a conveniently executed activity. Each of the steam generators is a horizontal pressure vessel with built-in design features to conveniently access and plug tubes. 6. Demineralized Water The primary coolant (which can also be referred to as the reactor coolant) is demineralized water, which promotes criticality safety because of its strong negative reactivity gradient with rise in temperature. Elimination of borated water also simplifies the nuclear steam supply system (NSSS) by eliminating the systems and equipment needed to maintain and control boron levels in the primary coolant. Pure water and corrosion resistant primary coolant loop help minimize crud buildup in the reactor pressure vessel. 7. Modularity One can build only one of the inventive nuclear reactor systems at a site, or a large number thereof. Clustering a number of inventive nuclear reactor systems at one site will reduce the overall O&M costs. 8. Long Operating Cycle The inventive nuclear reactor system will operate for approximately 3.5 years before requiring refueling. 9. Short Construction Life Cycle Virtually all components of the inventive nuclear reactor system are shop fabricated. Site work is limited to reinforced concrete construction and a limited amount of welding to assemble the shop-built equipment and parts. As a result, it is possible to complete the construction of one of the inventive nuclear reactor systems in 24 months from the first shovel in the ground. 10. Efficient Steam Cycle A pair of two horizontal steam generators are arranged in series and integrally welded to the reactor pressure vessel. The efficiency of the power cycle of the inventive nuclear reactor system, and its compactness, is further enhanced by superheaters that are integrally welded to the horizontal steam generators. The superheaters, one attached to each steam generator, increases cycle efficiency and also protect both the high pressure and low pressure turbines from the deleterious effect of moist steam. 11. Integral Pressurizer The design of the reactor pressure vessel incorporates an integral pressurizer that occupies the upper reaches of the reactor pressure vessel. The pressurizer serves to control the pressure in the reactor vessel. 12. Suitable for Water-Challenged Sites The inventive nuclear reactor system can be installed at sites with limited water availability, such as creeks and small rivers that are inadequate for large reactors. The inventive nuclear reactor system can be operated equally well in a water-challenged region by using air-cooled condenser technology to reject the plant's waste heat. Using air in lieu of water, of course, results in a moderate increase in the plant's cost. 12. System Parameters in the Safe and Proven Range The operating pressure and temperature within the reactor pressure vessel is in the proven range for PWRs. Lower core power density than that used in large PWRs for improved thermal-hydraulic control (please see table below) and an improved margin to departure-from-nucleate boiling in the reactor core. Exemplary System ParametersDataNumber of fuel assemblies in the core32Nominal thermal power, MWt446Nominal recirculation rate, MLb per hour5.46Reactor water outlet temperature, deg. F.580Reactor water inlet temperature, deg. F.333Reactor pressure, pounds per sq. inch2.250Water in the RV cavity, gallons30.00 13. Minimized Piping Runs and Minimum Use of Active Components to Enhance Reliability and Cost Competitiveness The amount of piping in the close-loop primary coolant circuit and the secondary coolant circuit in the inventive nuclear reactor system is the least of any nuclear plant design on the market, as is the number of pumps and valves. 14. In-Service Inspection All weld seams in the primary system including those in the reactor pressure vessel wall are available at all times for inspection. In particular, the weld seams in the reactor pressure vessel can be inspected by operating a manipulator equipped in-service inspection device in the reactor well during power generation. Thus, inventive nuclear reactor system exceeds the in-service inspection capability expected of nuclear plants under ASME Code Section XI. 15. Earthquake Hardened Design Virtually all major equipment in the inventive nuclear reactor system are either underground or horizontally mounted to withstand strong seismic motions. This includes the reactor pressure vessel, the fuel pool, the reactor water storage tank (all underground) and the horizontal steam generators, the horizontal superheaters, and the horizontal kettle reboiler that are floor mounted. 16. Aircraft Impact Proof Containment The containment structure of the inventive nuclear react system is designed to withstand the impact of a crashing fighter plane or a commercial liner without sustaining a thru-wall breach. II. Detail Referring now to FIG. 1, a natural circulation nuclear reactor system 1000 (hereinafter the “reactor system 1000”) is illustrated according to one embodiment of the present invention. The reactor system 1000 generally comprises a reactor pressure vessel 100 and a heat exchange sub-system 200. The reactor pressure vessel 100 contains a primary coolant 101 that is used to cool the rector core 102 and to heat a secondary coolant within the heat exchange sub-system 200. The reactor pressure vessel 100 is fluidly coupled to an incoming hot leg 201 of the heat exchange sub-system 200 via a primary coolant outlet port 103. Similarly, the reactor pressure vessel 100 is also fluidly coupled to an outgoing cool leg 202 of the heat exchange sub-system 200 via a primary coolant inlet port 104. As a result, a closed-loop primary coolant circuit 300 is formed through which the primary coolant 101 flows in a single-phase. As discussed in greater detail below, the flow of the primary coolant 101 through the closed-loop primary coolant circuit is a natural circulation flow induced by the heat given off by the normal operation of the reactor core 102. In certain embodiments, the internal cavity 105 of the reactor pressure vessel 100 is maintained under sufficient pressure to maintain the primary coolant 101 in a liquid-phase despite the high temperature within the rector pressure vessel 100. In the exemplified embodiment, a pressure control sub-system 50 (commonly referred to in the art as a pressurizer) is located within a top region of the reactor pressure vessel 100 and is configured to control the pressure of the internal cavity 105 of the reactor pressure vessel 100. The pressure control sub-system 50 is integral with the removable head 106 of the reactor pressure vessel 100 to prevent line break concerns and to provide a more compact reactor system 1000. Pressurizers are well known in the art and any standard pressurizer could be used as the pressure control sub-system 50. In one embodiment, the internal cavity 105 of the reactor pressure vessel 100 is maintained at a pressure in a range of 2000 psia to 2500 psia. In one more specific embodiment, the internal cavity 105 of the reactor pressure vessel 100 is maintained at a pressure between 2200 psia to 2300 psia. Of course, the exact pressure maintained in the internal cavity 105 of the reactor pressure vessel 100 is not to be limiting of the invention unless specifically claimed. The reactor pressure vessel 100 is an elongated tubular pressure vessel formed by a thick wall made of an acceptable ASME material, such as stainless steel. The reactor pressure vessel 100 extends from a bottom end 107 to a top end 108 along a substantially vertical axis A-A, thereby defining an axial length of the reactor pressure vessel 100. In one embodiment, the reactor pressure vessel 100 has an axial length of over 100 feet to facilitate an adequate level of turbulence in the recirculating primary coolant 101 from the natural circulation (also referred to as thermosiphon action in the art). In certain other embodiments, the reactor pressure vessel 100 has an axial length in a range between 100 feet to 150 feet. Of course, the invention is not so limited in certain alternate embodiments. The reactor pressure vessel 100 generally comprises a domed head 106 and a body 109. The domed head 106 is detachably coupled to a top end of the body 109 so as to be removable therefrom for refueling and maintenance. The domed head 106 can be coupled to the body 109 through the use of any suitable fastener, including bolts, clamps, or the like. In the exemplified embodiment, the body 109 comprises an upper flange 110 and the domed head 106 comprises a lower flange 111 that provided mating structures through which bolts 114 (FIG. 4) extend to couple the domed head 106 to the body 109. When the domed had 106 is coupled to the body 109, a hermetic seal is formed therebetween via the use of a gasket or other suitably contoured interface. The body 109 of the reactor pressure vessel 100 comprises an upstanding tubular wall 112 and a domed bottom 113 that hermetically seals the bottom end 107 of the reactor pressure vessel 100. The tubular wall 112 has a circular transverse cross-sectional profile in the illustrated embodiment but can take on other shapes as desired. In the exemplified embodiment, the domed bottom 113 is integral and unitary with respect to the tubular wall 112. Of course, in other embodiments, the domed bottom 113 may be a separate structure that is secured to the tubular wall 112 via a welding or other hermetic connection technique, such as the flanged technique described above for the domed head 106 and the body 109. Integral and unitary construct of the domed bottom 113 and the body 109 is, however, preferable in certain embodiments as it eliminates seams and/or interfaces that could present rupture potential. The reactor pressure vessel 100 forms an internal cavity 105 in which a reactor core 102 is housed. The reactor core 102 comprises nuclear fuel, in the form of fuel assemblies, as is known in the art. The details of the structure of the reactor core 102 are not limiting of the present invention in and the reactor system 1000 can utilize any type of reactor core or nuclear fuel. The reactor core 102 is positioned in a bottom portion 115 of the reactor pressure vessel 100. In one embodiment, the reactor core 102 has a core thermal power of 400 MWt to 600 MWt during the operation thereof. In one embodiment, the reactor core 102 is comprised of vertically arrayed fuel assemblies. The spacing between the fuel assemblies is governed by the design objective of keeping the reactivity (neutron multiplication factor) at 1.0 at all locations in the reactor pressure vessel 100. The criticality control in the axial direction is provided by the built-in neutron poison in the fuel rods (called IFBAs by Westinghouse) and possibly by control rods. A partition 120 is provided within the internal cavity 105 of the reactor pressure vessel 100 that divides the internal cavity into a primary coolant riser passageway 105A and a primary coolant downcomer passageway 105B. Both the passageways 105A, 105B are axially extending vertical passageways that form part of the closed-loop primary coolant circuit 300. In the exemplified embodiment, the partition 120 comprises an upstanding tubular wall portion 120A and a transverse wall portion 120B. The tubular wall portion 120A is an annular tube that is mounted within the internal cavity 105 of the reactor pressure vessel 100 so as to be concentrically arranged with respect to the upstanding wall 112 of the reactor pressure vessel 100. As a result, the primary coolant downcomer passageway 105B is an annular passageway that circumferentially surrounds the primary coolant riser passageway 105A. The primary coolant downcomer passageway 105B is formed between an outer surface 121 of the upstanding tubular wall portion 120A of the partition 120 and the inner surface 116 of the upstanding wall 112 of the reactor pressure vessel 100. The primary coolant riser passageway 105B is formed by the inner surface 122 of the upstanding tubular wall portion 120A of the partition 120. The transverse wall portion 120B is an annular ring-like plate that is connected to a top end of the of the upstanding tubular wall portion 120A of the partition 120 at one end and to the upstanding wall 112 of the reactor pressure vessel 100 on the other end. The transverse wall portion 120B acts a separator element that prohibits cross-flow of the primary coolant 101 between the primary coolant riser passageway 105A and the primary coolant downcomer passageway 105B within the top portion 117 of the reactor pressure vessel 100. In essence, the transverse wall portion 120B forms a roof of the primary coolant downcomer passageway 105B that prevents the heated primary coolant 101 that exits the reactor pressure vessel 100 via the primary coolant outlet port 103 from mixing with the cooled primary coolant 101 that enters the reactor pressure vessel 100 via the primary coolant inlet port 104, and vice-versa. Cross-flow of the primary coolant 101 between the primary coolant riser passageway 105A and the primary coolant downcomer passageway 105B is prohibited by the upstanding tubular wall portion 120A of the partition 120. In addition to physically separating the flow of the heated and cooled primary coolant 101 within the primary coolant downcomer and riser passageways 105A, 105B as discussed above, the partition 120 also thermally insulates the cooled primary coolant 101 within the primary coolant downcomer passageway 105B from the heated primary coolant 101 within the primary coolant riser passageway 105A. Stated simply, one does not want heat to transfer freely through the partition 120. Thus, it is preferred that the partition 120 be an insulating partition in the sense that its effective coefficient of thermal conductivity (measured radially from the primary coolant riser passageway 105A to the primary coolant downcomer passageway 105B) is less than the coefficient of thermal conductivity of the primary coolant 101. Making the effective coefficient of thermal conductivity of the partition 120 less than the coefficient of thermal conductivity of the primary coolant 101 ensures that the primary coolant 101 in the primary coolant downcomer passageway 105B remains cooler than the primary coolant 101 in the primary riser passageway 105A, thereby maximizing the natural circulation rate of the primary coolant 101 through the closed-loop primary coolant circuit 300. In a very simple construction, this can be achieved by creating the partition 120 out of a single solid material that has a low coefficient of thermal conductivity. However, it must be considered that the material should neither degrade nor deform under the operating temperatures and pressures of the reactor pressure vessel 100. In such an embodiment, the effective coefficient of thermal conductivity is simply the coefficient of thermal conductivity of the single solid material. In the exemplified embodiment, the low coefficient of thermal conductivity of the partition 120 is achieved by making the partition 120 as a multi-layer construction. As exemplified, the partition 120 comprises an insulating layer 124 that is sandwiched between two outer layers 125A, 125B. In one embodiment, the insulating layer 124 is a refractory material while the outer layers 125A, 125B are stainless steel or another corrosion resistant material. In certain embodiments, the insulating layer 124 is full encased in the outer layers 125A, 125B. The internal cavity 115 of the reactor pressure vessel 100 also comprises a plenum 118 at the bottom portion 115 of the reactor pressure vessel 100 that allows cross-flow of the primary coolant 101 from the primary coolant downcomer passageway 105B to the primary coolant riser passageway 105A. In the exemplified embodiment, the plenum 118 is created by the fact that the bottom end 123 of the upstanding tubular wall portion 120A of the partition 120 is spaced from the inner surface 119 of the domed bottom 113, thereby creating an open passageway. In alternate embodiments, the partition 120 may extend all the way to the inner surface 119 of the domed bottom 113. In such embodiments, the plenum 118 will be formed by providing a plurality of apertures/openings in the partition 120 so as to allow the desired cross-flow. The internal cavity 105 further comprises a plenum 126 at the top portion 117 of the reactor pressure vessel 100. The plenum 126 allows the heated primary coolant 101 that is rising within the primary coolant riser passageway 105A to gather in the top portion 117 of the reactor pressure vessel 100 and then flow transversely outward from the vertical axis A-A and through the primary coolant outlet port 103. The reactor core 102 is located within the primary coolant riser passageway 105A above the bottom plenum 118. During operation of the reactor core 102, thermal energy produced by the reactor core 102 is transferred into the primary coolant 101 in the primary coolant riser passageway 105A adjacent the reactor core 102, thereby becoming heated. This heated primary reactor coolant 101 rises upward within the primary coolant riser passageway 105A due to its decreased density. This heated primary coolant 101 gather in the top plenum 126 and exits the reactor pressure vessel 100 via the primary coolant outlet port 103 where it enters the heat exchange sub-system 200 as the incoming hot leg 201. In one embodiment, the heated primary coolant 101 entering the hot leg 201 of the heat exchanger has a temperature of at least 570° F., and in another embodiment a temperature in a range of 570° F. to 620° F. This heated primary coolant 101 passes through the heat exchange sub-system 200 where its thermal energy is transferred to a secondary coolant (described below in greater detail with respect to FIG. 2), thereby becoming cooled and exiting the heat exchange sub-system 200 via the cold leg 202. When exiting the cold leg 202 of the heat exchange sub-system, this cooled primary coolant 101 has a temperature in a range of 300° F. to 400° F. in one embodiment. In another embodiment, the heat exchange sub-system 200 is designed so that the temperature differential between the heated primary coolant in the hot leg 201 and the cooled primary coolant in the cold leg is at least 220° F. The cooled primary coolant 101 exiting the cold leg of the heat exchange sub-system 200 then enters the reactor pressure vessel 100 via the primary coolant inlet port 104, thereby flowing into a top portion 127 of the primary coolant downcomer passageway 105B. Once inside the primary coolant downcomer passageway 105B, the cooled primary coolant 101 (which has a greater density than the heated primary coolant 101 in the primary coolant riser passageway 105A) flows downward through the primary coolant downcomer passageway 105B into the bottom plenum 118 where it is drawn back up into the primary coolant riser passageway 105A and heated again by the reactor core 102, thereby completing a cycle through the closed-loop primary circuit 300. As discussed above, operation of the reactor core 102 causes natural circulation of the primary coolant 101 through the closed-circuit primary coolant circuit 300 by creating a riser water column within the primary coolant riser passageway 105A and a downcomer water column within the primary coolant downcomer passageway 105B. In one embodiment, the riser water column and the downcomer water column have a vertical height in a range of 80 ft. to 150 ft., and more preferably from 80 ft. to 120 ft. The vigorousness of the natural circulation (or thermosiphon flow) is determined by the height of the two water columns (fixed by the reactor design), and the difference between the bulk temperature of the two water columns (in water the SES and the downcomer space). For example, water at 2200 psia and 580° F. has density of 44.6 lb/cubic feet. This density increases to 60.5 lb/cubic feet if the temperature reduces to 250° F. The hot and cold water columns 60 feet high will generate a pressure head of 6.6 psi which is available to drive natural circulation of the primary coolant 101 through the closed-loop primary coolant circuit 300. A 90 feet high column will generate 50% greater head (i.e., 9.9 psi). As a result of the natural circulation of the primary coolant 101 achieved by the water columns and gravity, the reactor system 1000 is free of active equipment, such as pumps or fans, for forcing circulation of the primary coolant through the closed-loop primary coolant circuit. In the embodiment illustrated in FIG. 1, the primary coolant outlet port 103 is at a slightly lower elevation (1-3 ft.) than the primary coolant inlet port 104. However, in other embodiments, the primary coolant outlet port 103 and the primary coolant inlet port 104 will be at substantially the same elevation (see FIGS. 4 and 5). When the primary coolant outlet port 103 and the primary coolant inlet port 104 are at substantially the same elevation the partition 120 will be appropriately designed. Furthermore, as used herein, the term port includes mere apertures or openings. In one embodiment, the primary coolant 101 is a liquid that has a negative reactivity coefficient. Thus, the chain reaction in the reactor core 102 would stop automatically if the heat rejection path to the heat exchange sub-system 200 is lost in a hypothetical scenario. Thus, the reactor system 1000 is inherently safe. In one specific embodiment, the primary coolant 101 is demineralized water. All systems and controls used to maintain boron concentration in the reactor vessel in a typical PWR are eliminated from the reactor system 1000. Moreover, the use of demineralized water as the primary coolant 101 and the existence of the corrosion resistant surfaces of the reactor pressure vessel 100 help maintain crud buildup to a minimum. The reactivity control in the reactor core 102 is maintained by a set of control elements (burnable poisons) that are suspended vertically and occupy strategic locations in and around the fuel assemblies to homogenize and control the neutron flux. Referring now to FIGS. 1, 4 and 5 concurrently, it can be seen that a major portion 130 of the axial length of the reactor pressure vessel 100 located below a ground level 400 while a minor portion 131 of the axial length of the reactor pressure vessel 100 extends above the ground level 400. As such, the reactor core 102 is located deep below the ground level 400 while the heat exchange sub-system 200 is located above the ground level 400. In one embodiment, the heat exchange sub-system 200 is at an elevation that is 80 ft. to 150 ft, and preferably 80 ft. to 120 ft., greater than the elevation of the reactor core 102. The minor portion 131 of the reactor pressure vessel 100 includes a top portion 132 of the body 109 and the domed head 106. The primary coolant outlet port 103 and the primary coolant inlet port 104 are located on the minor portion 131 of the reactor pressure vessel 100 that is above the ground level 400. More specifically, the primary coolant outlet port 103 and the primary coolant inlet port 104 are located on the top portion 132 of the body 109 of the reactor pressure vessel 100 that is above the ground level 400. The major portion 130 includes a majority of the body 109 and the domed bottom 113. In certain embodiment, the major portion 130 of the reactor pressure vessel 130 is at least 75% of the axial length of the reactor pressure vessel 100. In other embodiments, the major portion 130 of the reactor pressure vessel 130 is between 60% to 95% of the axial length of the reactor pressure vessel 100. In another embodiment, the major portion 130 of the reactor pressure vessel 130 is between 75% to 95% of the axial length of the reactor pressure vessel 100. The reactor pressure vessel 100 comprises a reactor flange assembly 150 comprising a first reactor flange 151 and a second reactor flange 153. The top portion 132 of the body 109 of the reactor pressure vessel 100 is welded to the reactor flange assembly 150, which is a massive upper forging. The reactor flange assembly 150 also provides the location for the primary coolant inlet port 104 and the primary coolant outlet port 103 (FIGS. 4 and 5), and the connections to the heat exchange sub-system 200 (and for the engineered safety systems to deal with various postulated accident scenarios). This reactor flange assembly 150 contains vertical welded lugs 152 to support the weight of the reactor pressure vessel 100 in the reactor well 410 in a vertically oriented cantilevered manner (FIGS. 1 and 4). As a result, the reactor pressure vessel 100 is spaced from the wall surfaces 411 and floor surface 412 of the reactor well 410, thereby allowing the reactor pressure vessel 100 to radially and axially expand as the reactor core 102 heats up during operation and causes thermal expansion of the reactor pressure vessel 100. Furthermore, the major portion 130 of the reactor pressure vessel 100 is free of penetrations. In other words, the major portion 130 of the reactor pressure vessel 100 comprises no apertures, holes, opening or other penetrations that are either open or to which pipes or other conduits are attached. All penetrations (such as the primary coolant inlet and outlet ports 103, 104) in the reactor pressure vessel 100 are located in the above-ground minor portion 131, and more specifically in the top portion 132 of the body 109 of the reactor pressure vessel 100. In one embodiment, it is further preferred that the major portion 130 be a unitary construct with no connections, joints, or welds. The bottom portion 115 of the reactor pressure vessel 100 is laterally restrained by a lateral seismic restraint system 160 that spans the space between the body 109 of the reactor pressure vessel 100 and the wall surfaces 411 of the reactor well 410 to withstand seismic events. The seismic restraint system 160, which comprises a plurality of resiliently compressible struts 161, allows for free axial and diametral thermal expansion of the reactor vessel. The bottom of the reactor well 410 contains engineered features to flood it with water to provide defense-in-depth against a (hypothetical, non-mechanistic) accident that produces a rapid rise in the enthalpy of the reactor's contents. Because the reactor system 1000 is designed to prevent loss of the primary coolant 101 by leaks or breaks and the reactor well 410 can be flooded at will, burn-through of the reactor pressure vessel 100 by molten fuel (corium) can be ruled out as a credible postulate. This inherently safe aspect simplifies the design and analysis of the reactor system 1000. Referring now to FIGS. 2 and 4-5 concurrently, an embodiment of the heat exchange sub-system 200 is illustrated. While a specific embodiment of the heat exchange sub-system 200 will be described herein, it is to be understood that, in alternate embodiments, one or more of components can be omitted as desired. For example, in certain embodiments, one or both of the horizontal superheaters 205, 206 may be omitted. In certain other embodiments, one of the horizontal steam generators 203, 204 may be omitted and/or combined into the other one of the horizontal steam generators 203,204. Moreover, additional equipment may be incorporated as necessary so long as the natural circulation of the primary coolant 101 through the closed-loop primary coolant circuit 300 is not prohibited through the introduction of substantial head loss. As mentioned above, the heat exchange subsystem 200 comprises an incoming hot leg 201 that introduces heated primary coolant into the portion of the closed-loop primary coolant circuit 300 that passes through the heat exchange sub-system 200 and an outgoing cold leg 202 that removes cooled primary coolant from the portion of the closed-loop primary coolant circuit 300 that passes through the heat exchange sub-system 200. In order to minimize (and in some embodiments eliminate) pressure loss in the closed-loop primary coolant circuit 300 caused by an increase in the elevation of the primary coolant flow, the steam generators 203, 204 and the superheaters 205, 206 are all of the horizontal genre (i.e., the tubes which carry the primary coolant extend substantially horizontal through the shell-side fluid) and are in horizontal alignment with each other where possible. Within the heat exchange sub-system 200, the primary coolant flow of the closed-loop primary coolant circuit 300 is divided into two paths 211, 212 at a flow divider 215. The flow divider 210 can be a three-way valve, a three-way mass flow controller, or a simple Y plumbing joint. The first path 211, which carries the majority of the primary coolant flow, travels through the first horizontal steam generator 203 and then through the second horizontal steam generator 204. Meanwhile, the second path 212, which carries a minority of the primary coolant flow, travels through the first horizontal superheater 205 and then through the second horizontal superheater 206. After passing through the first and second horizontal steam generators 203, 204 and the first and second horizontal superheaters 205, 206, the first and second paths 211, 212 converge in a flow converger 216, which combines the primary coolant flows of the first and second paths 211, 212 and directs the combined flow to the outgoing cold leg 202. As with the flow divider 215, the flow converger 216 may be a three-way valve, a three-way mass flow controller, or a simple Y plumbing joint. In one embodiment, 10% to 15% of the incoming primary coolant flow that enters the heat exchange sub-system 200 via the hot leg 201 is directed into the second path 212 while the remaining 85% to 90% of the incoming primary coolant is directed into the first path 211. In one specific example, the incoming primary coolant that enters the heat exchange sub-system 200 via the hot leg 201 has a flow rate of 5 to 7 million lbs./hr. In this example, 0.6 to 1 million lbs./hr. of the primary coolant is directed into the second path 212 while the remainder of the primary coolant flow is directed into the first path 211. The first and second horizontal steam generators 203, 204 are operbaly coupled in series to one another along the first path 211 of the closed-loop primary coolant circuit 300. Both of the horizontal steam generators 203, 204 are horizontally disposed shell-and-tube heat exchangers. The first horizontal steam generator 203 is a high pressure steam generator while the second horizontal steam generator 204 is a low pressure steam generator (in comparison to the high pressure steam generator). The high first steam generator 203 is located upstream of the second horizontal steam generator 204 along the closed-loop primary coolant circuit 300. Similarly, the first and second horizontal superheaters 205, 206 are operbaly coupled in series to one another along the second path 212 of the closed-loop primary coolant circuit 300. The first horizontal superheater 205 is a high pressure superheater while the second horizontal superheater 206 is a low pressure superheater (in comparison to the high pressure superheater). The high first steam superheater 205 is located upstream of the second horizontal superheater 206 along the closed-loop primary coolant circuit 300. Furthermore, the first and second superheaters 205, 206 are located in parallel to the first and second horizontal steam generators 203, 204 along the closed-loop primary coolant circuit 300. Furthermore, the first and second horizontal steam generators 203, 204 are interconnected by a return header so that the hot primary coolant entering the first horizontal steam generator 203 heats the secondary coolant to make steam for the high-pressure turbine 220 and then proceeds to the second horizontal steam generator 204 with minimal pressure loss to make steam for the low-pressure turbine 221. The flow of the primary coolant in the first path 211 is used to convert a secondary coolant flowing through the shell-side of the first and second horizontal steam generators 203, 204 from liquid-phase to gas-phase through the transfer of heat form the primary coolant to the secondary coolant within the first and second horizontal steam generators 203, 204. Because the flow of the primary coolant through the first and horizontal second steam generators 203, 204 is substantially horizontal in nature, the flow of the primary coolant through the first path 211 does not cause any substantial pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation. Moreover, because of the horizontal alignment of the first and second horizontal steam generators 203, 204 with each other and the primary coolant outlet and inlet ports 103, 104 of the reactor pressure vessel 100 (FIG. 5), the primary coolant flow that travels along the first path 211 from the primary coolant outlet port 103 of the reactor pressure vessel 100 to the primary coolant inlet port 104 of the reactor pressure vessel 100 does not cause any substantial pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation. While the achievement of substantial zero pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation is exemplified in terms of a horizontal flow, it is possible that such substantial zero pressure drop can be achieved by a decline in elevation as the primary coolant flows downstream in the closed-loop primary coolant circuit 300. The flow of the primary coolant in the second path 212 is used to superheat the vapor-phase of the secondary coolant exiting the first and second horizontal steam generators 203, 204 via the first and second horizontal superheaters 205, 206 respectively, thereby further drying the vapor-phase of the secondary coolant. The use if the horizontal superheaters enhance the thermodynamic efficiency of the turbine cycle, carried out on the high pressure turbine 220 and the low pressure turbine 221. The first and second horizontal superheaters 205, 206 are horizontally disposed shell-and-tube heat exchanger positioned directly above (and in series) with the first and second steam generators 203, 204 (FIG. 5). However, due to the slight increase in the elevation of the superheaters 205, 206 resulting from their location above the first and second horizontal steam generators 203, 204, the flow of the primary coolant in the second path 212 does cause some pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation. However, because only a small amount (10% to 15%) of the total primary coolant that flows through the heat exchange subsystem 200 is directed into the second path 212 and through the horizontal superheaters 205, 206, the pressure drop does not significantly affect the desired natural circulation. Moreover, the increase in elevation is negligible when compared to the height of the flow driving water columns. In such an embodiment, at least 85% of the flow of the primary coolant through the heat exchange sub-system 200 is still entirely horizontal from the primary coolant outlet 103 to the primary coolant inlet 104 and does not cause any substantial pressure drop in the closed-loop primary coolant circuit 300 due to increase in elevation. Further, in certain alternate embodiments, the horizontal superheaters 205, 206 could be eliminated and/or repositioned to be in horizontal alignment with the horizontal steam generators 203, 204. As shown in FIG. 5, the first and second horizontal steam generators 203, 204 are coupled directly to the each other and to the reactor pressure vessel 100. More specifically, the inlet of the first horizontal steam generator 203 is coupled directly to the primary coolant outlet port 103 of the reactor pressure vessel 100 while the outlet of the first horizontal steam generator 203 is coupled directly to the inlet of the second horizontal steam generator 204. The outlet of the second horizontal steam generator 204, is in turn, coupled directly to the primary coolant inlet port 104 of the reactor pressure vessel 100. The first and second horizontal steam generators 203, 204 are arranged so as to extend substantially parallel to one another, thereby collectively forming a generally U-shaped structure. Thus, the first path 211 also takes on a generally U-shape In certain embodiments, the first and second horizontal steam generators 203, 204 are integrally welded to the reactor vessel 100 and to each other. Referring now to FIGS. 2 and 3A-B, each of the first and second horizontal steam generators 203, 204 comprise a preheating zone 208, 210 and a boiling zone 207, 209. Both of the first and second horizontal steam generators 203, 204 are of the single-pass type in which the primary coolant flow of the first path 211 is the tube-side fluid. Each of the single-pass tubes 330 extend substantially horizontally through the preheating zones 208, 210 and the boiling zones 207, 209. The secondary coolant circuit has a main feedwater intake 501 and a return to condenser exit 502 into and out of the heat exchange sub-system 200 respectively. The secondary coolant, which is in the liquid-phase 505, enters each of the first and second horizontal steam generators 203, 204 along line 503. The incoming liquid phase 505 of the secondary coolant is preheated within the preheater zones 208, 210 of the first and second horizontal steam generators 203, 204. The secondary coolant in liquid-phase 505 flows through a tortuous path as shell-side fluid in the preheater zones 208, 210 and then enters the boiling zones 207, 209, where it is further heated by the primary coolant flow passing through the tubes 330. In the boiling zones 207, 209, the liquid-phase secondary coolant 505 vaporizes and exits the first and second horizontal steam generators 203, 204 as high pressure and low pressure steam 504 that is respectively supplied to the high and low pressure turbines 220, 221. The shells of the horizontal steam generators 203, 204 and the horizontal superheaters 205, 206 provide additional barriers against potential large-break LOCAs, as do the turning plenum and the eccentric flanges that join the steam generators 203, 204 to the reactor pressure vessel 100, as shown in FIGS. 4 and 5. All systems connected to the reactor vessel 100 use a similar approach to ensure that there is no potential for a large-break LOCA that could rapidly drain the water from the reactor vessel 100 and uncover the reactor core 102. As long as the reactor core 102 is covered under all potential conditions of operation and hypothetical accident, the release of radioactive material to the public is minimal. As explained in the foregoing, the reactor system 1000 is an intrinsically safe reactor which, in the event of a problem external to the reactor containment building or within containment, is designed to automatically shut down in a safe mode with natural circulation cooling. Nevertheless, to instill maximum confidence, a number of redundant safety systems can be engineered to protect public health and safety under hypothetical accident scenarios that are unknown or unknowable, i.e., cannot be mechanistically postulated. In the case of an abnormal condition when the normal heat transport path through the steam generators are not available, then the pressure in the reactor vessel 100 will begin to increase. In such a case rupture discs will breach allowing the reactor coolant to flow into a kettle reboiler located overhead. The kettle will have a large inventory of water that will serve to extract the heat from the reactor coolant until the system shuts down. Diverse systems perform duplicate or overlapping functions using different physical principles and equipment to ensure that a common-mode failure is impossible. As used throughout, ranges are used as shorthand for describing each and every value that is within the range. Any value within the range can be selected as the terminus of the range. In addition, all references cited herein are hereby incorporated by referenced in their entireties. In the event of a conflict in a definition in the present disclosure and that of a cited reference, the present disclosure controls. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims. |
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abstract | A storage container is carried into a nuclear-reactor containment, and then, the core structure is housed in the storage container. In this process, a bottom cover is carried to a storage-container body by using a moving device. The storage-container body and the bottom cover are positioned by adjusting the position of the bottom cover on the moving device while the bottom cover is supported by free bearings. The bottom cover is then attached to the storage-container body. |
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045483477 | summary | This disclosure relates to the assembly of nuclear fuel assemblies, particularly for use in fast breeder reactors. The present state of the art involves fabrication of fuel assemblies by manual manipulation or by mechanical processes incapable of limiting spread of radioactive contamination to adjoining components and equipment. In such configurations, the fabrication steps are time consuming and create significant radiation contamination control problems. To achieve flexibility of process steps, radiation contamination control, and efficient fabrication in an apparatus having production capability, the present invention has been directed toward development of a self-contained automated loading system. It allows for system variation and contamination control without degrading production capability. More specifically, the present apparatus permits fabrication of fuel assemblies by batch processing methods, using equipment which can be physically and environmentally isolated as required. U.S. Pat. No. 4,167,959 exemplifies an apparatus in which a manual glove box is provided for filling fuel rods. A seal is provided at the entrance of the glove box for engaging the cladding. Contamination is stated to be controlled by differences in pressure between an antechamber, a processing chamber, and the exterior of the glove box. The use of a glove box for processing of nuclear fuel assemblies is carried further in the disclosure of U.S. Pat. No. 4,070,240. The patent discloses an automated transport conveyor for moving fuel elements between successive assembly stations. A common sealing apparatus is stated to prevent external contamination as the fuel elements are advanced between the various manual stations. A large scale system for manufacturing nuclear fuel pellets is disclosed in U.S. Pat. No. 4,174,938. The system includes process components arranged vertically and providing for gravity flow of the product from one component to the next. The various process components are modular and each can be removed without interfering with the others. Physical isolation of the components is provided by appropriate seals and manual access is accomplished through glove ports. Another large scale system for cleaning nuclear fuel elements is shown in U.S. Pat. No. 4,063,962. Batches of fuel elements are suspended vertically and moved through the components of the system. Airlocks and seals are provided to contain contamination. A number of prior patents have been directed specifically to sub-systems for loading fuel pellets into nuclear fuel elements by automated or semi-automated equipment. Representative disclosures are found in U.S. Pat. Nos. 3,746,190, 4,125,577, 3,907,123 and 4,158,601. U.S. Pat. No. 3,711,993 relates to an airlock or cylindrical seal for engagement about the periphery of fuel cladding. U.S. Pat. No. 3,828,518 discloses a welding apparatus for closing the end of a fuel rod by use of a rotating electrode head. SUMMARY OF THE INVENTION It is an object of this invention to provide a novel conveyor and support system for handling a batch of fuel pins during fabrication processes. The system permits batch processing of the fuel pins in combination with individual fabrication steps so as to more efficiently adapt automated fuel pin fabrication processes to a production schedule. More specifically, the system lends itself to batch inerting of the fuel pins in an apparatus interposed between subsystems which individually fill and cap the fuel pins. Another object of this invention is to provide a system in which the various work stations and subsystems are modular. They can be duplicated when required by production quotas, and individual components can be removed or substituted as necessitated by servicing and repair schedules. Another object of the invention is to provide an integrated system for fabricating fuel pins which can be totally automated, thereby removing the necessity of using manual glove boxes during fuel pin fabrication. Another object of the invention is to provide a unique system of handling individual fuel pins by which each fuel pin can be reciprocated axially and rotated about its axis. This eliminates the necessity of rotating work elements used to carry out particular fabrication steps, such as the welding or the cleaning of fuel pin surfaces. Another object of the invention is to provide a unique system of seals so as to assure effective protection against contamination of vulnerable mechanical components and to permit efficient inerting of the fuel pin interiors. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects and in accordance with the purposes of the present invention as embodied and broadly described herein, the apparatus of this invention may comprise conveyor means for moving parallel fuel pin assemblies along a path perpendicular to their lengths, transport means for coaxially positioning a fuel pin assembly along a preselected operational axis, transfer means for shifting individual fuel pin assemblies between the conveyor means and the transport means, powered means for shifting the transport means relative to the conveyor means in a direction parallel to the operational axis, and fuel pin handling means for receiving one end of a fuel pin assembly shifted thereto by the powered means. By this apparatus, fuel pin assemblies can be moved to and from the transport means by the conveyor means and can be shifted axially for operations carried out by the fuel pin handling means. The apparatus provides an effective automated system for moving a plurality of fuel pin assemblies between automated production stations including fuel pin handling means by which the fuel pins can be filled, cleaned, inerted, capped and welded. All of these steps can be effectively accomplished without manual intervention. |
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050769936 | summary | BACKGROUND OF THE INVENTION The present invention relates to nuclear-based contraband detection systems, and more particularly to an apparatus and method for accurately detecting contraband concealed within a container, such as a suitcase, truck or other object. As used herein, the term "contraband" includes, but is not limited to, explosives and illicit drugs. Diagnostic nuclear techniques in general involve use of two highly penetrating radiations (e.g., neutrons and gamma rays) which enable one to detect concealed explosives or other contraband materials. The radiations act as follows: An appropriately fashioned primary radiation excites atomic nuclei within a designated volume of an object. The excited atomic nuclei subsequently relax, emitting electromagnetic or particle radiation in the process that is characteristic of the nuclear species. The analysis of the emitted spectrum thus facilitates the detection of a particular substance within the object, e.g., explosives or illegal drugs. In other words, if the emitted spectrum includes radiation of a given energy, then the presence of a particular element within the object can be inferred. Thus, a spectrum showing characteristic radiation lines of particular intensities serves as a "signature" that identifies the presence of a particular chemical element within the object being examined. Identifying the chemical elements and/or chemical compounds within an object thus involves identifying the corresponding signatures that are present in the radiations emitted from the material. See e.g., Gozani, Active Nondestructive Assay of Nuclear Materials, United States Nuclear Regulatory Commission, NUREG-CR-0602, SAI-FM-2585 (1981). It is common practice to use neutrons as the primary radiation and to measure the ensuing gamma-ray spectra for non-intrusive diagnostic purposes. U.S. Pat. No. 3,832,545 and patent application Ser. No. 07/053,950, filed 05/26/87, for example, disclose nuclear-based explosive detection systems that make use of neutrons of mainly thermal energies. In contrast, European Patent publication EP-O-227-497-A1 discloses a nuclear-based explosive detection system wherein fast neutrons of energies from 7 to 15 million electron volts (MeV) are employed. Disadvantageously, the thermal neutron based detection systems provide, for practical purposes, primarily only one signature of the four cardinal constituents of explosives, namely the signature of nitrogen (and possibly hydrogen). The fast neutron based detection system, on the other hand, may provide signatures of all four ingredients of explosives, or other contraband, thus enhancing the interrogating power of the fast neutron contraband detection systems. (The four cardinal chemical constituents of explosives are hydrogen, carbon, nitrogen, and oxygen.) It must be observed, however, that simply obtaining the signatures of the constituent elements of a specified contraband does not necessarily indicate that such contraband is present in the object under investigation. This is because many benign materials (non-contraband) also include such elements. A great diagnostic advantage may thus be obtained when a three-dimensional image of the distribution of element densities within the interrogated body is also formed, as such image of densities may help further distinguish contraband from non-contraband. A suitable three-dimensional image for this purpose may advantageously be obtained by performing a section-by-section neutron irradiation of the object, and by performing a computer-based analysis of the energy and intensity of the gamma rays that are produced in each section. Such analysis has in the past required the judicious positioning of gamma-ray detectors around the object, as taught in Applicants' earlier patent application, Ser. No. 07/053,950, filed 05/26/87. As previously indicated, neutron interrogation of objects for the detection of contraband, e.g., explosives, is known in the art. One of the most common forms of neutron interrogation, and the only form that has yet been commercialized, is thermal neutron activation (TNA). In the TNA techniques, the object being interrogated is exposed to low energy neutrons, causing gamma rays having an energy characteristic of the element(s) within the object to be generated. The gamma rays of a particular energy are detected and counted. From such count, a determination can be made as to the abundance of nitrogen within the object being interrogated. The ability of TNA techniques to reliably detect the explosives depends greatly on the large nitrogen content and density of the explosive. Another technique known in the art for detecting explosives is fast neutron activation (FNA). FNA techniques are similar to TNA techniques in that an object being interrogated is bombarded with neutrons. However, in the case of FNA, the neutrons have a higher energy, e.g., 14 MeV, and the gamma rays they generate allow the presence of additional elements to be detected. In particular, FNA allows the presence of hydrogen, carbon, and oxygen to be detected in addition to nitrogen. The relative concentrations of all of these elements thus comprise a "signature" that further helps to identify a particular substance, i.e., contraband. A still further technique for detection of explosives involves detection of the alpha particle generated in a T(d,n).sup.4 He reaction which produces a 14 MeV neutron. The neutron and alpha particle are emitted in opposite directions. A small particle detector near the tritium target detects the alpha particle. The corresponding neutron is emitted at 180.degree. within a solid angle equal to the solid angle subtended by the alpha detector from the target. This solid angle defines a "beam" of neutrons that is used to interrogate a sample, such as a suitcase or other container. A gamma ray detector is placed near the sample, detecting gamma rays in coincidence with the alpha particles. The time difference between the alpha particle detection and gamma ray detection can provide the position of the gamma ray source along the beam. By scanning the beam, a three-dimensional image of the gamma ray sources can thus be generated. Finally, as indicated in French patent document #EP 0 227 497 A1, and a paper recently presented in the 5th Asia/Pac Aviation Seminar in Kuala Lumpur (Aug. 17-21, 1987), it is also known in the art to combine both fast and thermalized neutrons in the same detection system. As indicated in these documents, a partially moderated 14 MeV pulsed neutron source is used with one or more well shielded germanium detector(s). Nitrogen and oxygen are determined through (n,x.gamma.) reactions during the bursts of the fast neutrons, and hydrogen and chlorine are determined between pulses through (n,.gamma.) reactions with thermalized neutrons. SUMMARY OF THE INVENTION In general and simple terms, the present invention provides a highly effective and direct manner for using neutrons to "look inside" a closed object or container, such as a cargo truck or a piece of luggage, to determine the abundance of particular atomic nuclei, i.e., elements, therewithin. Once such abundance is known, the identification of particular contraband is readily accomplished, as all contraband of interest contains specific atomic elements in identifiable proportions and densities. As with prior art approaches using nuclear techniques, the "looking inside" of the object is achieved by detecting gamma rays produced in nuclear reactions. The gamma rays have energies characteristic of the particular atomic nuclei, which nuclei represent the residuals of these nuclear processes. Unlike prior art approaches, in which the object is immersed in a bath of thermal (low energy) neutrons, thereby causing a large number of gamma ray sources to be produced in an extended volume, and thereby necessitating the use of a large number of gamma ray detectors and a rather complex processing procedure to reconstruct useful image data from the gamma ray data, the present invention uses a highly collimated short pulse of fast (high energy) neutrons to sequentially interrogate small volume elements ("voxels") of the object. In this manner, the present detection system is thus able to "see" right into a particular voxel and directly determine what elements are present therein based on the gamma rays that are detected. By looking in a controlled (and rapid) sequence into a sufficient number of voxels in this manner, a direct indication is thus obtained of the abundance of prescribed chemical elements within the object. In keeping with one aspect of the present invention, a rapid yet effective system for the reliable detection of designated types of contraband, including explosives, using pulsed fast neutron activation (PFNA), is provided. In one embodiment, the pulsed beams of fast neutrons are collimated to a very high degree, i.e., pencil beams, using the kinematics of a A(B,n)-nuclear endothermic reaction where M.sub.B >>M.sub.A. The pulse width is on the order of a nanosecond (preferably less). These pulsed neutrons are directed to the object under investigation and cause (n,x.gamma.) reactions (preferably with x.ident.n') in a limited small object volume that is defined by the intersection of the pencil beam and the screened object. By choosing appropriately the lateral section of the scanning neutron pencil beam, i.e., the reaction kinematics, if required augmented with an external collimation, and by measuring the time-of-flight of the interacting neutrons, a convenient subdivision of the object into a string of small volume elements, i.e., "voxels", is realized. By precisely controlling the time of occurrence including duration of the neutron burst and determining the time of gamma ray detection, it is thus possible, to determine the particular region, or voxel, from which the gamma ray is produced. Since the highly penetrating fast neutrons have a high probability for gamma ray production nuclear reactions with the elements oxygen, carbon, chlorine and nitrogen, the carbon, nitrogen, oxygen, and chlorine content of a particular voxel can advantageously be determined directly and precisely, which determination leads directly to an indication as to whether such particular voxel contains contraband. By combining this information from a substantial sample of the voxels that make up the object, the presence (or absence) of any type of contraband within the object can be inferred quantitatively with a high degree of reliability. In addition, by relaxing the fast timing condition, other chemical elements, such as hydrogen, can be determined advantageously through the neutron thermalization process and its subsequent radiative capture in the screened object. In an alternative embodiment of the present invention, the pulsed fast neutrons are produced in the A(B,n) reactions where M.sub.B .ltoreq.M.sub.A, i.e., reactions such as D(d,n), T(d,n) or Li(p,n). Subdivision of the interrogated object into voxels using this embodiment is obtained by using an external neutron beam collimator and may involve the positioning of gamma ray detectors. In a still further alternative embodiment, the interrogated object is irradiated from many directions using, e.g., multiple sources of pulsed fast neutrons, including appropriate transport of the ion beams and/or movement and/or rotation of the object. The sizes and loci of the voxels are derived in accordance with this embodiment mainly from the measurements of the time of flight of the neutrons and gamma rays. The present invention may thus be characterized as a contraband detection apparatus that includes: (1) scanning means for scanning an object under investigation with a pulsed beam of fast neutrons; (2) first detecting means for detecting gamma rays emitted from the object as a result of interactions between a neutron from the pulsed beam of fast neutrons and an atomic nucleus within the object; (3) identifying means for identifying the particular atomic element which gives rise to the detected gamma ray; (4) locating means for determining the approximate location within the object of the origin of each gamma ray detected by the detecting means; and (5) second detection means responsive to the scanning means, identifying means and locating means for detecting a distribution and concentration of at least one atomic element within the object indicative of the presence of contraband. Another embodiment of the invention may be characterized as a contraband detection system comprising: (1) means for generating a recurring short pulse of directed fast neutrons; (2) means for scanning an object under investigation for the presence of contraband with this recurring short pulse of fast neutrons, each of the fast neutrons possibly reacting with a particular atomic nucleus present within the object, thereby generating gamma rays having an energy characteristic of the particular atomic nucleus with which the fast neutrons react; (3) means for detecting the gamma rays produced by neutrons in reactions with atomic nuclei, the detection means including means for detecting the energy of a particular gamma ray and the time of its detection relative to the time of generation of the short pulse of fast neutrons; and (4) means for determining a particular volume element, or voxel, within the object from which a particular detected gamma ray originated, the detected gamma ray thereby providing a direct indication of the particular atomic nuclei in the particular voxel. Using this system, the particular atomic nuclei present in a sample of the voxels within the object advantageously provide a direct indication of the abundances and distributions of particular elements within the object. This determination, in turn, provides a direct indication of the presence or absence of contraband, as the presence of contraband is indicated by a prescribed abundance and distribution of the particular elements within the object. Further, the present invention may be characterized as a system for detecting contraband comprising: (a) means for scanning an object under investigation with a pulsed beam of fast neutrons by controllably directing the pulsed beam at a prescribed volume of the object; (b) means for detecting gamma rays having prescribed energies emitted from the prescribed volume of the object as a result of interactions between the pulsed fast neutrons and atomic nuclei of particular elements within the prescribed volume, the prescribed energies corresponding to atomic elements commonly found in contraband; and (c) means for ascertaining whether a distribution and concentration of at least one atomic element indicative of contraband exists within the prescribed volume. Using this system, the determination that contraband is present within the prescribed volume advantageously allows investigation of the object using pulsed fast neutron to be terminated, thereby reducing the amount of time required by the system to detect contraband within the object. That is, if contraband is found in a single voxel of the object, there is no need to continue searching through other voxels of the object. However, as desired and/or required, additional prescribed volumes, e.g. adjacent voxels, of the object can be similarly investigated in order to confirm the presence of contraband within the object. The invention also includes a method of detecting contraband comprising the steps of: (a) directing a pulsed beam of fast neutrons towards a prescribed volume of an object under investigation; (b) detecting gamma rays having prescribed energies emitted from the prescribed volume of the object as a result of interactions between pulsed fast neutrons and atomic nuclei of particular elements within the prescribed volume of the object, the prescribed energies corresponding to atomic elements commonly found in contraband; and (c) repeating steps (a) and (b) for a sufficiently large number of small prescribed volumes of the object under investigation so as to ascertain whether a distribution and concentration of at least one atomic element indicative of contraband exists within the object. It is a feature of the present invention to provide a contraband detection system that has improved sensitivity, i.e., an improved ability to reliably detect the elements that make up prescribed contraband, regardless of the form of the contraband or the manner in which the contraband may be arranged or hidden within the object. As a result of this improved sensitivity, the detection system of the present invention advantageously provides a higher probability of detection (PD) and a lower Probability of False Alarm (PFA) than has heretofore been possible with prior art contraband detection systems. It is another feature of the present invention to provide such a detection system wherein the requisite information upon which a contraband/non-contraband decision is based is obtained directly from scanning data obtained from the object under investigation without significant additional processing, and/or probabilistic assessments, thereby allowing a noticeable improvement in the throughput time, i.e., the time it takes to put an object through the system to determine if it contains prescribed contraband. It is still another feature of the invention to provide a contraband detection system that is available for use with more diverse types and forms of objects to be examined. For example, where the object being examined is luggage, the present invention may examine all sizes and types of luggage, from small carry-on parcels, to larger check-in luggage, to full luggage carts and other large containers. Where the object being examined is a cargo truck, all sizes of trucks or equivalent cargo-carrying vehicles can be readily examined using the system of the present invention. Another feature of the invention allows the detection of contraband to occur without using a large number of gamma-ray detectors, as has heretofore been required with nuclear-based detection systems. With prior art systems, such as is disclosed in U.S. Patent Ser. No. 07/053,950, it has been necessary to surround the object being examined with a large number of detectors so that the particular detector that sensed an emitted gamma ray could also provide some indication as to the location within the object from where the gamma-ray originated, thereby helping to form a density map of the contents of the object. With such prior art approach, it is necessary to not only determine that a gamma ray of a specified energy has been detected, but it is also necessary to know and track the particular detector (within an array of a large number of detectors) where the detection occurred. In contrast, the present invention only requires a few detectors because it is the time-of-flight measurement and the lateral size of the neutron beam that determines the origin of the gamma ray of a particular energy (and hence the location of a particular element), and this measurement is not significantly dependent upon the location of the detector. As a result, the present invention advantageously provides an image of the elemental distribution within the object directly without the need for a large detector array or complex mathematical reconstruction. Still further, it is a feature of the present invention to provide a detection system that exhibits an improved signal-to-noise ratio. This improved signal-to-noise ratio results largely from the pulsed nature of the highly collimated interrogating neutrons. That is, the collimated pulsed neutrons produce gamma rays only in desired regions or voxels of the object during a prescribed time window. Hence, the amount of background noise (i.e., gamma rays not of interest, such as those produced in the detector or in other locations or regions of the object or its surrounding environment) present in the time windows of interest is significantly reduced, thereby improving the signal-to-noise ratio. A further feature of the invention provides for the detection of contraband without the necessity of detecting associated particles, which associated particle detection is mandatory in some prior art systems. Advantageously, because associated particles, e.g., alpha particles, do not need to be detected, the intensity of the interrogating beam is not limited as it is in systems where associated particles are detected (in which systems the count rate capability of the associated particle detection channel severely limits the beam intensity). It is yet another feature of the invention to provide a reliable nuclear-based contraband detection system employing fast (high energy) neutrons that exhibits significantly reduced shielding requirements, geometry constraints and equipment specifications over prior art nuclear-based detection systems. |
claims | 1. An electron beam apparatus for irradiating a sample with a primary electron beam, and detecting secondary electrons generated from a surface of the sample by the irradiation to evaluate the sample surface, comprising:a primary electro-optical system for focusing the primary electron beam on the sample surface and scanning it with the primary electron beam;a secondary electro-optical system comprising at least one stage of lens;an E×B separator for separating the secondary electrons generated from the sample surface and directing them to the secondary electro-optical system;a first detector for detecting the secondary electrons which have passed through the secondary electro-optical system;a second detector for detecting an exposure dose of the primary electron beam on the sample surface;a memory for storing the detected exposure doses on the sample surface; anda control apparatus adapted to calculate a unit exposure dose per unit area on the basis of the stored exposure doses, and controlling the primary electron beam such that the unit exposure dose does not exceed a predetermined level. 2. An electron beam apparatus according to claim 1, wherein the secondary electro-optical system further comprises a diaphragm. 3. An electron beam apparatus according to claim 1, wherein the primary electro-optical system comprises an aperture plate having a plurality of apertures by which a plurality of primary electron beams are formed from an electron beam emitted from an electron gun. 4. An electron beam apparatus according to claim 1, further comprising an apparatus for outputting a signal representing a position on the sample surface, at which the unit exposure dose exceeds the predetermined level. 5. An electron beam apparatus according to claim 1, whereinthe sample is a wafer;the electron beam apparatus further comprises an evaluation apparatus for evaluating the wafer surface on the basis of the detected secondary electrons, the evaluation being executed for every constant stripe width on the wafer while a stage carrying wafer is moving; andthe calculation of the unit exposure dose is executed for an area which is smaller than an area of (the stripe)×(a length of a chip in the stripe direction). 6. An electron beam apparatus according to claim 1, further comprising:a third detector for detecting a moving speed of a sample stage;a compensation apparatus included in at least one of the primary electro-optical system and the secondary electro-optical system, for compensating, in response to the moving speed of the sample stage detected by the third detector, a deflection amount of at least one of the primary electron beam and the secondary electron beam. 7. An electron beam apparatus according to claim 1, further comprising:a third detector for detecting a moving speed of a stage carrying the sample;a fourth detector for detecting a position of the stage;a compensation apparatus included in at least one of the primary electro-optical system and the secondary electro-optical system, for compensating, in response to the moving speed and the position of the sample stage detected by the third and fourth detectors, a deflection amount of at least one of the primary electron beam and the secondary electron beam. 8. An electron beam apparatus according to claim 1, whereinthe primary electron beam is irradiated on the sample surface in the form of a multi-beam; andthe minimum interval between adjacent primary electron beams on the sample surface is wider than the resolution of the secondary electro-optical system. 9. An electron beam apparatus according to claim 1, whereinthe primary electron beam is irradiated on the sample surface in the form of a multi-beam; andthe multi-beam is set such that when the beams on the sample surface are projected onto an axis perpendicular to a moving direction of a sample stage, the projected points on the axis are spaced at substantially the same interval. |
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abstract | A betavoltaic power source. The betavoltaic power source comprises a source of beta particles, a substrate with shaped features defined therein and a InGaP betavoltaic junction disposed between the source of beta particles and the substrate, and also having shaped features therein responsive to the shaped features in the substrate, the InGaP betavoltaic junction device for collecting the beta particles and for generating electron hole pairs responsive thereto. |
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description | An RPV is a huge apparatus, which comes to be about 25 m in height, about 6 m in diameter, and about 1000 tons in weight. The RPV is provided with a main steam outlet nozzle, a feedwater nozzle, etc., as well as 50 to 60 nozzles. A pipe is connected to each of those nozzles. When carrying out such an RPV from its containment building, therefore, pipes connected to those nozzles and steel structures disposed outside the RPV must be cut off and removed. Because the RPV has been exposed to radiation, those pipes must be cut off and removed in a highly radiation-exposed area. To avoid being exposed to radiation, therefore, measures must be taken for protecting the work from such radiation. Furthermore, in addition to those pipes, such structures as a fuel change bellows, a bulk head plate, etc. are connected to the RPV. To carry the RPV out from the containment building, therefore, those structures must also be cut off and removed. The total weight of those structures will become about 100 tons. Because those structures also include radioactive substances, they must be housed in containers (hereafter, to be referred to as casks) that can shield radioactive rays during the work for carrying out those structures from the containment building. To put those structures in casks, it is also required that those structures are cut off in accordance with the sizes of the casks. The load of such a work will become enormous. In addition, casks are expensive, since they are required to shield radioactive rays. This is why the background art method has confronted with a problem that many expensive casks have been required for housing those removed structures. Furthermore, because the surfaces of those casks must be inspected for contamination before the casks are handled in which structures are housed and before each cask is transported and stored independently, the work will also become enormous in load. Consequently, the background art has been confronted with another problem that the work period becomes long. Furthermore, a pressurized water reactor has confronted with such problems when not only the containment vessel, but also the steam generator are carried out from the containment building. Hereunder, the embodiments of the present invention will be described with reference to the accompanying drawings. [First Embodiment] At first, a description will be made for a method for carrying out the equipment of a nuclear power plant in the first embodiment of the present invention. In this embodiment, the present invention applies to a method for carrying out the RPV of a boiling-water reactor (hereafter, to be referred to as a BWR). According to this embodiment, at first pipes connected to the nozzles of the RPV are cut off sequentially from top to bottom thereof while the water level in the reactor is lowered step by step. Then, the upper lattice plate is removed and placed vertically on the upper flange of the shroud in the reactor. After that, wastes are carried into the shroud through a gap between the RPV and the upper lattice plate, then the upper lattice plate is returned to its horizontal position on the top surface of the shroud. Then, other wastes are piled on each another on the upper lattice plate, so that the RPV are carried out together with those wastes from the containment building and stored in a storage. The wastes in the shroud are also carried out at this time, of course. FIG. 2 shows a schematic vertical cross sectional view of a BWR-4 containment building to which the carry-out procedure shown in FIG. 1(a) is applied in this first embodiment. In the containment building 31 is disposed an RPV 1. An incore structure 2 is housed in the RPV 1. In the containment building 31 are housed a PCV 16 in the lower part of an operation floor 80 and in the PCV 16 is housed an RPV 1. The RPV 1 is disposed on an RPV pedestal 18 disposed inside the PCV 16. Outside the RPV 1 is disposed a reactor shielding wall 17 (hereafter, to be referred to an RSW) used to shield the radioactive rays from the RPV 1. In the upper part of the PCV 16 are disposed a reactor well 32 for pooling water when the fuel assembly is replaced with another and/or structures inside the RPV 1 are taken out; a spent fuel pool 33 for storing spent fuel assemblies; an equipment pool 81 for placing such incore-structures as a steam drier removed for periodical inspection, etc. Inside the spent fuel pool 33 is disposed a fuel rack 33a for storing spent fuel assemblies. On top of the RPV 1 is disposed a reactor pressure vessel head 37 (hereafter, to be referred to as an RPV head). As shown in FIG. 3, the RPV 1 is disposed in the center of the PCV 16 and the RPV 1 is fixed to the RPV pedestal 18 with foundation bolts 28. On the side wall of the RPV 1 are disposed a main steam nozzle 9 for feeding the steam generated in the RPV 1 to the power generator; a feedwater nozzle 10 for feeding condensate water into the RPV 1; a reactor spray nozzle 11 for cooling the inside of the RPV 1; inlet and outlet nozzles 12 and 13 of a recirculation system for circulating reactor water in the RPV 1; a reactor core measuring nozzle 14 for various instrumentations in the RPV 1; a reactor core differential pressure instrumental nozzle 56 for measuring the pressure in the reactor; and an RPV drain nozzle 57 for draining reactor water from the RPV 1. An insulating material 192 is disposed on the outer periphery of the RPV 1. In the upper part of the PCV 16 are disposed a fuel change bellows 15 for partitioning the inner space of the PCV 16 and a bulk head plate 19. Inside the RPV pedestal 18 are disposed a control rod housing 23 (hereafter, to be referred to as a CRD housing) for guiding the control rods; a CRD housing support beam 22 (hereafter, to be referred to as a CRD beam) for supporting the CRD housing 23; a CRD housing support block 25 (hereafter, to be referred to as a CRD block); and an ICM housing 24 for supporting a monitor of neutrons in the reactor (hereafter, to be referred to as an ICM). The RSW 17 is fixed to the RPV pedestal 18 with foundation bolts 29. In the upper part of the RSW 17 are disposed a PCV stabilizer 30, which is an anti-vibration supporting member of the PCV 16 and an RPV stabilizer 30a, which is an anti-vibration supporting member of the RPV 1. As shown in FIG. 4, the reactor core shroud (hereafter, to be referred to as a shroud) is disposed in the center of the reactor. The shroud 5 is supported by a shroud support cylinder 82. The shroud support cylinder 82 is supported by a baffle plate 84 and a shroud support leg 83 at the bottom of the RPV 1. Inside the shroud 5 is disposed an upper lattice plate 7, which supports the upper reactor core in its upper part and a reactor core supporting plate 6, which supports the lower reactor core in its lower part. In addition, control rods 85, control rod guide tubes 86, and fuel assemblies 87 are disposed in the shroud 11. A jet pump 8 is disposed between the shroud 5 and the RPV 1. In the upper part of the shroud 5 are disposed a steam drier 3; a steam separator and shroud head assembly 4; a guide rod 88, a feedwater sparger and pipe assembly 89, and a reactor core spray sparger and pipe assembly 90. The steam separator and shroud head 4 and the shroud 5 are fastened with bolts via ribs (not illustrated) disposed in the upper part of the shroud 5. In the lower part of the shroud 5 are disposed such equipment as an ICM housing 24, incore stabilizer 27, etc. The RPV head 37 is fixed to the RPV 1 with stud bolts 37a attached to the flange 37b. At the bottom of the RPV 1 are disposed a CRD housing 23 for storing a control rod driver 20 (hereafter, to be referred to as a CRD) and an ICM housing 24 for the ICM 21. In the RPV 1 are disposed various types of equipment inside the reactor such way. The equipment in the reactor is divided into an incore structure 2 installed inside the RPV 1 and other structures than the incore one. The incore structure 2 is composed of a steam drier 3, a steam separator and shroud head 4, a shroud 5, a reactor core supporting plate 6, an upper lattice plate 7, a jet pump 8, etc. Structures other than the in-core structure 2 are a main steam nozzle 9, a feedwater nozzle 10, a reactor core spray nozzle 11, a recirculation system inlet nozzle 12, a recirculation system outlet nozzle 13, a reactor core measuring nozzle 14, a reactor core differential pressure instrumental nozzle 56, an RPV drain nozzle 57, a CRD housing 23, and an ICM housing 24. Those structures are all disposed on the side wall (shell) of the RPV 1. In this embodiment, the RPV is carried out from its nuclear power plant composed as described above. The RPV is carried out in accordance with the flowchart shown in FIG. 1(a). Hereunder, a method for carrying out such an RPV in this embodiment will be described. At first, a periodical inspection is done for the object nuclear power plant by disassembling the generator in step S1. In step S2, the reactor is opened. At this time, water is filled in the reactor well 32, that is, up to the level 67a of the reactor water 67. The step for opening the reactor includes removing of the PCV 16, removing of the RPV head 37, removing of the RPV head, removing of the steam drier 3, and removing of the steam separator and shroud head 4. The steam drier 3 and the steam separator and shroud head 4 are, when they are removed, moved into the equipment pool 81. When the steam drier 3 and the steam separator and shroud head 4 are removed, a large space is made at the upper side of the upper lattice plate 7 in the RPV 1. The work for opening the reactor is a necessary critical one required to remove the fuel assemblies 87 from the reactor. In this embodiment, the steam drier 3 and the steam separator and shroud head assembly 4 that are removed are used again as are. They may be replaced with new ones as needed. Next, all the fuel assemblies 87 are removed from the RPV 1 in step S3 (FIG. 5(a)). The removed fuel assemblies 87 are then moved into a fuel rack 33a disposed in the spent fuel pool 33. At this time, the reactor water level 67a is set so as to fill the reactor well 6. Next, control rods 85 and control rod tubes 86 are carried out in step S4. The control rods 85 and the control rod tubes 86 are thus all moved into the fuel rack 33a disposed in the spent fuel pool 33. In this embodiment, it is premised that the control rods 85 and the control rod tubes that are removed are used again as they are. However, they may be replaced with new ones as needed. Consequently, a large space is made inside the reactor core shroud 5, since the fuel assemblies 87, the control rods 85, and the control rod tubes 86 are all removed from the reactor core shroud 5. The processes in steps S1 to S4 are the same as those in a normal periodical inspection. Next, the upper lattice plate 7 is removed from the top of the reactor core shroud 5 and placed vertically in the upper part of the reactor core shroud 5 (FIG. 5(b)). The upper lattice plate 7 is thus placed vertically as shown in (FIG. 5(c)) in the upper part of the reactor core shroud 5 after it is removed from the top thereof. At this time, the reactor water level 67a must be set so as to fill the reactor well 32, since the radiation dose of the upper lattice plate 7 is high. Because the upper lattice plate 7 placed vertically is larger than the diameter of the reactor core shroud 5, the plate 7 is caught by the upper flange of the reactor core shroud 5. Consequently, it never falls in the reactor core shroud 5. The upper lattice plate 7 placed vertically is then fixed with a wire, etc. The upper lattice plate 7 placed vertically can thus be prevented from falling. At this time, a path can be secured at the bottom side of the upper lattice plate 7 inside the RPV 1 for wastes to be carried out from the containment building while the upper lattice plate 7 is left in the RPV 1. In addition, because the upper lattice plate 7 is left in the RPV 1 such way, the radiation dose to the surroundings can be reduced more than when the plate 7 is taken out from the RPV 1. Next, processes in step S6 and step S20 are done in parallel. In step S6, the structures surrounding the RPV are disassembled. In step S20, the structures in the RPV pedestal 18 are disassembled. Although the processes in steps S6 and S20 are done together in this embodiment, the process in step S20 may be done after the process in step S6 or vice versa. Next the process in step S6 will be described. In this step S6, the RPV 1 is disconnected completely from the structures surrounding the RPV 1. In this embodiment, joints enclosed by a dotted line in FIG. 5(a) are disconnected. The details of the process in step S6 are steps S7 to S12. Hereunder, those detailed processes from step S7 to step S13 will be described sequentially. In step S7, the reactor water level 67a is lowered up to the RPV flange 1c as shown in FIG. 6. Then, the bulk head plate 19 and the fuel change bellows 15 are removed. The bulk head plate 19 and the fuel change bellows 15, when they are removed, are placed on the reactor core supporting plate 6 in the shroud 5 by inserting them through a gap between the vertically placed upper lattice plate 7 and the RPV 1. After that, the PCV stabilizer 30 is removed in step S8, then it is placed on the reactor core supporting plate 6 in the shroud 5 by inserting it through the gap between the vertically placed upper lattice plate 7 and the RPV 1. When this PCV stabilizer 30 is removed, a path is secured for moving the structures positioned between the RPV 1 and the PCV 16 and under the PCV stabilizer 30 into the RPV 1 via the reactor well 32. Next, step S10 will be described. As shown in FIG. 1(b), steps S101 to S109 are the details of step S10. At first, the reactor water level 67a is kept at the top of the RPV flange 1c as shown in FIG. 7 in step S101. Then, a water sealing plug 51 is inserted in the reactor so as to close the nozzle 9. Next, the main steam pipe 9a connected to the main steam nozzle 9 is cut off at a joint with the main steam nozzle 9 and at a bent-down portion of the main steam pipe 9a. After that, a nozzle sealing plate 52 is attached to the main steam nozzle 9 from outside the RPV 1. Then, cut-off pieces of the pipe 9a are placed on the reactor core supporting plate 6 in the shroud 5 by inserting them through the gap between the vertically placed upper lattice plate 7 and the RPV 1. The main steam pipe 9a may also be cut off by a method that the reactor water level is lowered under the opening of the main steam nozzle 9, then the reactor water level 67a is returned to the top of the RPV flange 1c after the nozzle sealing plate 52 is attached to the main steam nozzle 9. Next, both feedwater pipe 10a and reactor core spray pipe 11a connected to the feedwater nozzle 10 and the reactor core spray nozzle 11 are cut off in step S102, then they are moved into the RPV 1. Because the feedwater nozzle 10 and the reactor core spray nozzle 11 cut off in this step are connected to a feedwater sparger 10b and an incore spray pipe 11b, the water sealing plug 51 cannot be inserted in each of those nozzles 10 and 11, although it is possible in step S101. Consequently, the reactor water level 67a is lowered below the opening of the feedwater nozzle 10 and the opening of the reactor core spray nozzle 11 and the feedwater pipe 10a outside the RPV 1 and the reactor core spray pipe 11a are cut off respectively. Then, the nozzle sealing plate 52 is attached to both of the feedwater nozzle 10 and the reactor core spray nozzle 11 respectively. FIG. 8(a) shows a state of the feedwater pipe 10a cut off from the feedwater nozzle 10. The feedwater pipe 10a is cut off at a joint with the feedwater nozzle 10 and at a bent-down portion of the feedwater pipe 10a respectively after the reactor water level 67a is lowered below the opening of the feedwater nozzle 10. After the cutting, a nozzle sealing plate 52 is attached to the feedwater nozzle 10 from outside the RPV 1. Then, the cut-off pieces are placed on the reactor core supporting plate 6 in the shroud 5 by inserting them through the gap between the vertically placed upper lattice plate 7 and the RPV 1. The reactor core spray nozzle 11, as well as the reactor core spray pipes 11a and 11b are cut off in the same way. Finally, the reactor water level 67a is returned to the position of the RPV flange 1c. At this time, when the feedwater pipe 10a and the reactor core spray pipe 11a are cut off, the reactor water level 67a is kept at the position set in step S101 as shown in FIG. 8(b) with respect to the feedwater nozzle 10, the feedwater pipe 10a, and the feedwater sparger 10b. Then, the feedwater sparger 10b connected to the feedwater nozzle and/or incore spray pipe 11b connected to the reactor core spray nozzle 11 may be cut off from inside the RPV. After that, a water sealing plug 51 is inserted in each nozzle from inside the RPV, and then the feedwater pipe 10a and the reactor core spray pipe 11a may be cut off from outside the RPV and a nozzle sealing plate 52 may be attached to each of those pipes 10a and 11a in the same procedure as that in step S101. In this case, because the nozzle sealing plate 52 can be attached to each object pipe without lowering the reactor water level 67a, the reactor water will shield and further reduce the radiation dose of the workers. The following procedure may also be used in step S102. At first, as shown in FIG. 8(a), the reactor water level 67a is set lower than the opening of the feedwater nozzle 10 and the opening of the reactor core spray nozzle 11, then the feedwater sparger 10b connected to the feedwater nozzle 10 and/or incore spray pipe 11b connected to the reactor core spray nozzle 11 are cut off from inside the RPV, then the water sealing plug 51 is inserted in each of those nozzles from inside the RPV. Next, the reactor water level 67a is raised up to the position set in step S101. After that, the feedwater pipe 10a is cut off at a junction with the feedwater nozzle 10 and at a bent-down portion of the feedwater pipe 10a. In the same way, the core spray pipe 11a is cut off. After that cutting, the nozzle sealing plate 52 is attached to the feedwater nozzle 10 and the reactor core spray nozzle 11 from outside the RPV respectively. Then, cut-off pipe pieces are placed on the reactor core supporting plate 6 in the shroud 5 by inserting them through the gap between the upper lattice plate 7 and the RPV 1. If this procedure is to be employed, it is possible to cut off the feedwater sparger and the reactor core spray sparger outside the reactor water. The cutting will thus become more easily. In addition, when the feedwater pipe 10a and the core spray pipe 11a are cut off, the reactor water can shield the radioactive rays, thereby the radiation dose of the workers can be reduced. Next, step S103 will be described. In this step S103, the upper lattice plate 7, which has been placed vertically, is returned to its horizontal position as shown in FIG. 9. The reactor water level 67a is set at the position of the RPV flange 1c at this time. In this example, it is after the process in step S102 that the upper lattice plate 7 is returned to its horizontal position. However, the plate 7 may be returned to its horizontal position in step S103 if any trouble is expected to occur in the process for returning the plate 7 to its horizontal position due to extra structures to be carried into the shroud in step S101 or S102. It is also possible to move the upper lattice plate 7 into the equipment pool without placing the plate 7 vertically in step S5. The upper lattice plate 7 may be returned in the upper part of the shroud 5 in step S103 after object structures are all carried in the pool. After the process in step S103, structures to be carried into the RPV 1 are placed on the upper lattice plate 7 (???). In this case, it is no need to fix the upper lattice plate 7 with wire, etc., so the work is simplified and done easily. Next, step S104 will be described. In this S104 step, the core measuring pipe 14a connected to the core measuring nozzle 14 is cut off and moved into the RPV 1. The core measuring nozzle 14 is disposed close to the fuel while the reactor is operating, so the radiation dose is high around the core measuring nozzle 14. Therefore, if the RSW plug (door) 17a is released and the core measuring pipe 14a outside the RPV 1 is cut off after the reactor water 67 is lowered up to the lower side of the core measuring nozzle 14, then the radiation dose of the workers might exceed a regulated value in the process for cutting off the pipe 14a. To avoid this, therefore, the following method is employed. At first, the reactor water level 67a is set lower than the core measuring nozzle 14 as shown in FIG. 22(a). Then, the pipe 14a is cut off at the position 95 set outside the RSW plug 17a. After that, as shown in FIG. 22(b), the water sealing plug 51 is inserted in the opening of the core measuring pipe 14a made by the cutting, thereby sealing the core measuring nozzle 14 from water. Then, as shown in FIG. 22(c), the reactor water level 67a is raised up to the upper side of the shroud 5. Next, the RSW plug 17a is released and the core measuring pipe 14a is cut off at a joint with the core measuring nozzle 14. After that, a nozzle sealing plate 52 is welded to the outer side of the core measuring nozzle 14. The plate 52 may be fastened with bolts. Then, the RSW plug 17a is closed. After the setting, cut-off pipe pieces are carried into the RPV 1 and placed on the upper lattice plate 7(???). Consequently, the core measuring nozzle 14 protruded from the RPV 1 through the RSW 17 can be cut off while the dose of the workers is suppressed within a regulated value. The core measuring pipe 14a may also be cut off not only at a joint between the core measuring pipe 14a and the core measuring nozzle 14, but also at a position protruded from the RPV 1 when a water sealing plug 51 is attached if the protruded portion does not interfere with the inside of the RSW 17 when the RPV 1 is moved upward. Next, step S105 will be described. In step S105, the reactor water level 67a set in step S104 is kept as is while a water sealing plug 51 is attached to a jet pump 8 and the inlet pipe 12a of the recirculation system is cut off and carried into the reactor. As shown in FIG. 10, the inlet nozzle 12 of the recirculation system is connected to the jet pump 8 inside the reactor. Therefore, an inlet mixer 8a, which is part of the jet pump 8, is removed first, then a water sealing plug 51 is inserted in the removed portion. Then, the inlet pipe 12a of the recirculation system is cut off outside the RPV 1 at a joint with the inlet nozzle 12 of the recirculation system and at a bent-down portion of the inlet pipe 12a of the recirculation system. After the cutting, a nozzle sealing plate 52 is attached to the inlet nozzle 12 of the recirculation system outside the RPV 1. After that, the cut-off pipe pieces are placed on the upper lattice plate 7. With this procedure, the reactor water can shield the radioactive rays, so the dose of the workers can be reduced. This process may also be done as follows; when the inlet pipe 12a of the recirculation system is to be cut off, the reactor water level 67a is lowered up to the lower side of the opening of the inlet nozzle 12 of the recirculation system, then the inlet pipe 12a of the recirculation system is cut off and a nozzle sealing plate 52 is attached to the nozzle 12, then the reactor water level 67a is returned to the upper part of the RPV flange. In such a case, it is no need to remove the inlet mixer 8a. The work is thus simplified. Next, step S106 will be described. In step S106, the reactor water level 67a set in step S104 is kept as is while the outlet pipe 13a of the recirculation system is cut off and carried into the reactor. At first, as shown in FIG. 11, a water sealing plug 51 is inserted in the nozzle 13 at the outlet of the recirculation system from inside the reactor. Then, the outlet pipe 13 of the recirculation system is cut off from inside the RPV 1 at a joint with the outlet nozzle 13 of the recirculation system and at a bent-down portion of the outlet pipe 13 of the recirculation system. After the cutting, a nozzle sealing plate 52 is attached to the outlet nozzle of a recirculation system 13 from outside the RPV 1. The cut-off pipe pieces are then placed on the upper lattice plate 7. According to this procedure, the reactor water can shield the workers from radioactive rays, thereby the dose of the workers can be reduced. Next, step S107 will be described. In step S107, pipes connected to all the nozzles of the RPV, which are disposed lower than the outlet nozzle 13 of the recirculation system, are cut off and placed on the upper lattice plate 7. The core differential pressure instrumental nozzle 56 is also cut off at this time. When those pipes are cut off, the reactor water level 67a is set lower than the joint of each pipe with the RPV 1 before the nozzles are cut off. After each nozzle cutting, a nozzle sealing plate 52 is attached to the object nozzle and cut-off pieces are placed on the upper lattice plate 7. For the nozzles positioned lower than the outlet nozzle 13 of the recirculation system to be cut off in step S107, the reactor water shields the workers from radioactive rays, since they work away from the reactor area. The radiation dose will thus become comparatively low. The workers"" exposure to radiation will also be suppressed within a regulated value during the work for cutting off and carrying out such pipes as the core differential pressure instrumental pipe 56a connected to the core differential pressure instrumental nozzle 56. Next, step S108 will be described. In step S108, the RPV drain pipe 57a connected to the RPV drain nozzle 57 is cut off and a nozzle sealing plate 52 is attached to the nozzle 57 after the reactor water 67 is drained completely from the RPV 1 through the RPV drain nozzle 57. The RPV drain pipe 57a connected to the RPV drain nozzle 57 is disposed between the CRD housings 23, so the pipe 57a is carried out together with those CRD housings 23. The RPV drain pipe 57a is cut off at a position in the RPV pedestal 70 so that the pipe 57a, when carried together with the RPV 1, does not interfere with the RPV pedestal 70. Then, a nozzle sealing plate 52 is attached to the cut-off portion of the pipe 57a. Then, the cut-off pipe pieces are placed on the upper lattice plate 7 (FIG. 12). This competes the process in step S10. Next, step S11 will be described. In step S11, the RPV head 37 is placed on the RPV flange with use of a ceiling crane, then the head 37 is attached to the RPV 1 with RPV stud bolts 37a. Next, step S12 will be described. In step S12, the RPV stabilizer 30a, which is an anti-vibration supporting member of the RPV 1, is removed. Then, the RSW 17 is separated from the RPV 1. Next, step S20 will be described. In step S20, structures in the RPV pedestal are disassembled, then the RPV 1 is disconnected from the RPV pedestal 18. As shown in FIG. 1(a), step S20 is divided into steps S21 to S25. In step S21, the CRD block 25 is removed as follows. At first, the nuts fastening the CRD block are loosened, thereby the CRD block is removed. After that removal, the CRD block is stored in a storage area (not illustrated) outside the RPV pedestal. Next, the cables are disconnected from both CRD 20 and ICM 21 in step S22. For the ICM, the cable terminal connector is released first, then the cable is removed. For the CRD, control rods are disconnected from the CRD first, then the PIP (CRD position detector) connector is removed, thereby both cable and PIP are drawn out from the CRD. After that removal, the CRD block, as well as cables and PIP are moved into the storage area (not illustrated) prepared outside the RPV pedestal. Next, the CRD 20 is removed in step S23 as follows. At first, the bolts in the lower part of the CRD are loosened with use of a CRD changer (not illustrated) installed at the RPV pedestal, then the CRD itself is removed. The CRD is placed in the storage area (not illustrated) prepared outside the RPV pedestal. Processes in steps S21 to S23 are usually carried out in a periodical inspection. Next, a pipe 26 used to insert/draw out the CRD is cut off in step S24 as follows. The pipe 26 is cut off at a position around the inner wall of the RPV pedestal. After that, the CRD beam 22 is removed in step S25 after it is cut off with use of a cutting machine. A gas may be used to cut off the CRD beam 22 instead of the cutting machine used in steps S24 and S25. In this embodiment, processes in steps S6 and S20 can be carried out in parallel so as to shorten the working time. The process in step S20 may also be carried out after the process in step S6 or vice versa. Next, a crane 65 is installed in step S30, thereby an opening 31a for carrying out the RPV 1 is made in the roof of the containment building (hereafter, to be referred simply to as an opening) in step S40 as follows. As shown in FIG. 19, at first the crane 65 is installed outside the containment building 31, then an opening 31a is made in the roof of the containment building 31. In addition, opening equipment 64 is installed in the temporary opening 31a. The equipment 64 can open/close the opening 31a. The temporary opening 31a is made in the upper part of the RPV 1 in a size decided according to the size of the RPV to be carried in, the size of a shield disposed for the RPV 1, the size of a lifting jig for carrying out the shield, swinging of the shield when in carrying it in and out, etc. Because the size of the temporary opening 31a is decided by taking the swinging of each structure to be passed through this temporary opening 31a into consideration, these structures to be passed through the opening 31a can avoid coming in contact with the containment building. In addition, the negative pressure of the containment building 31 can be managed properly in rains and/or during a work with use of the opening equipment 64 to be opened and closed freely. The crane 65 may be installed any time in step S30 and the opening 31a may be made any time in step S40 if they are done before the shield is attached in step S50. Next, the cylindrical shield 60 is carried into the containment building in step S50 with use of the crane 65. Then, as shown in FIG. 13, the shield 60 is placed on the top surface of the RSW 17 temporarily. Then, a strong back (lifting jig) 63 is attached to the RPV 1. After that, the RPV 1 is lifted up and an RPV 1 stabilizer bracket 30b is fit in the stopper 61 attached to the upper part of the shield 60 as shown in FIG. 20. Consequently, the shield 60 can be attached in a reactor core area, in which fuel assemblies are loaded during a normal RPV operation. After that, a curing sheet 60a is put on the following structure in step S60. As shown in FIG. 21, both RPV 1 and shield 60 are lifted up by the crane 60, then the RPV head 37, the lower CRD housing 23, etc. that are positioned higher than the shield, are covered with the curing sheet 60a, then the end portions of the curing sheet 60a are sealed fixedly with sealing tape (not illustrated). The curing sheet 60a may be a vinyl chloride sheet, or the like. With this, the RPV 1 loaded with wastes and sealed there and to be carried out of the containment building are covered with both shield 60 and curing sheet 60a, thus shielded completely from radioactive rays. Next, the RPV 1 is lifted up in step S61 as follows. At first both RPV 1 and shield 60 are lifted up with use of the crane 65 as shown in FIG. 14, then carried out of the containment building 31 through the opening 31a. Before the carry-out operation, the shield 60 is checked for surface contamination. After the carry-out operation, the equipment 64 is closed. Next, both RPV 1 and shield 60 are placed in a storage in step S62 as follows. As shown in FIG. 15, a vertical underground type storage 66 is prepared around the containment building 31 and the crane 65 is turned towards the storage 66 while lifting up the shield 60 carried out of the containment building 31. The shield 60 is then carried into the storage 66. After the carry-in operation, the storage 66 is closed and sealed with a lid. Both RPV 1 and shield 60 may also be put on a trailer and transported to a storage prepared far away from the containment building. The storage 66 may be prepared in a building connected to the containment building 31. According to this embodiment, because wastes are put in an RPV and carried out together with the RPV from the object containment building, it is possible to carry out a plurality of wastes at once from the containment building. The number of times for carrying out the wastes can thus be reduced more than the background art method for carrying out RPV and wastes separately. Consequently, the working time for carrying out wastes from the containment building can also be reduced. In addition, the number of casks and shields used while those wastes go half around the containment building can be reduced. In addition, because the RPV is lifted up in a shield placed temporarily on the top surface of the RSW when the shield is attached to the RPV, the shield can be attached to the RPV easily. In addition, the radiation dose applied into the containment building from the reactor core area of the RPV can be reduced. (Second Embodiment) Next, a description will be made for a method for carrying out the equipment of a nuclear power plant in the second embodiment of the present invention. This embodiment describes a method that a hole is made in the upper lattice plate 7 installed on the top of the reactor core shroud 5 in step S5 in the first embodiment, then cut-off or removed pipes and wastes of structures are put in the reactor core shroud 5 through the hole. Hereunder, this embodiment will be described with reference to FIG. 1(c). Except for that step S5 in FIG. 1(a) in the first embodiment is replaced with step S5a shown in FIG. 1(c) and step S103 shown in FIG. 1(b) in the first embodiment is deleted in this embodiment, other procedures are the same as those in the first embodiment. The descriptions for the same items will thus be omitted, avoiding redundant description. Next, step S5a will be described. In step S5a, the lattice 7a of the upper lattice plate 7 located on top of the reactor core shroud 5 is cut off and a hole is made in the upper lattice plate 7 as shown in FIG. 16. The lattice 7a of the upper lattice plate 7 is about 15 mm in thickness. This lattice 7a can thus be cut off and a hole can be made in the upper lattice plate 7. In addition, cut-off chips of the lattice 7a generated when the hole is made in the upper lattice plate 7 are placed in the reactor core shroud 5. At this time, the reactor water level 67a must be set so as to fill the reactor well 32, since the radiation dose of the upper lattice plate 7 is high. This embodiment can thus obtain the same effect of the first embodiment. In addition, because a hole is made in the upper lattice plate 7, such wastes as pipes, etc. can be carried in the shroud 5 located in the lower part of the upper lattice plate 7 without removing the upper lattice plate 7 from the upper part of the shroud 5. In addition, because the upper lattice plate 7 is not removed, a work for fixing the plate 7 is omissible. (Third Embodiment) Next, a description will be made for a method for carrying out the equipment of a nuclear power plant in the third embodiment of the present invention. The method employed in this third embodiment enables cut-off and removed pipes and structures to be placed on the upper lattice plate 7 in the RPV 1. Processes in steps S5 and S103 in the first embodiment are deleted in this embodiment. In addition, steps S5 and S103 shown in FIG. 1(a) are deleted from the flowchart in this embodiment. Pipes and structures cut off and removed in step S6 are carried onto the upper side of the upper lattice plate 7 as shown in FIG. 17. Except for that steps S5 and S103 in the first embodiment are deleted from this embodiment and the destination of the items to be carried out in step S6 is different from that in the first embodiment, other items are the same as those in the first embodiment. Description for those same items will thus be omitted in this embodiment. In step S6, pipes and structures that are cut off and removed are placed on the upper lattice plate 7. If it is expected that any of those pipes and structures placed on the upper lattice plate 7 might fall into the shroud 5, an iron plate, etc. may be placed on the upper lattice plate 7 so as to prevent such the falling. This embodiment can thus obtain the same effect as that of the first embodiment. In addition, because the upper lattice plate 7 is not processed at all, the number of processes can be reduced. In addition, because the upper lattice plate 7 is not removed, a process for fixing the plate 7 can be omitted. Because the number of processes can be reduced, the working time can be reduced. If all of the pipes and structures that are cut off and removed cannot be placed in the RPV 1, surplus ones may be carried out from the containment building with use of another shielded container. (Fourth Embodiment) Next, a description will be made for a method for carrying out the equipment of a nuclear power plant in the fourth embodiment of the present invention. According to the method in this fourth embodiment, a water sealing plug made of an elastic material is inserted in each nozzle inside the RPV 1 and the pipe connected to the plugged nozzle is cut off. After that, a closing plate is welded to the nozzle from outside the RPV 1 or by another means during a process for cutting off and carrying out pipes in step S10 in the first embodiment. Other procedures are the same as those in the first embodiment, thus the description for them will be omitted here. A balloon, which is an elastic bag made of such an elastic material as rubber or the like, is used as the water sealing plug 51. Next, step S101 in this embodiment will be described. As shown in FIG. 18(a), the reactor water level 67a is kept in the upper part of the RPV flange 1c. Then, the balloon 54 made of an elastic material is inserted in each nozzle from inside the reactor. After that, a liquid hardening agent is fed into the balloon 54 so as to blow it up. Then, it is awaited until the fluid gardening agent is hardened. The balloon 54 functions as a water sealing plug 51 for closing the main steam nozzle 9. Next, the main steam pipe 9a connected to the main steam nozzle 9 is cut off at a joint with the main steam nozzle 9 and at a bent-down portion of the main steam pipe 9a. After the cutting, a closing plate 52 is attached to the nozzle 9 from outside the RPV 1 as shown in FIG. 18(b). After that, the cut-off pipe pieces are placed on the reactor core supporting plate 6 in the shroud 5 by inserting them through the gap between the vertically placed upper lattice plate 7 and the RPV 1. A sealing water process is done in the same procedure as the above so as to cut off the object pipes in steps S102 to S108. This completes the process in step S10. This embodiment can also obtain the same effect as that in the first embodiment. In addition, each object nozzle can be closed by sealing water with use of an elastic material, since the water sealing plug sticks fast to the nozzle. According to this embodiment, a pipe connected to each nozzle can be cut off and the nozzle can be plugged while the reactor water level is kept at the upper side of the opening of the nozzle when each nozzle is cut off in step S10. Consequently, the radiation dose of the workers can be reduced. The balloon 54 may be a water-absorbent-sealed one. Air and water are fed into the balloon so as to harden the balloon. In addition, the balloon 54 may have a gas check-valve (ex., air, nitrogen gas) and a gas or a gas and low density mortar are fed into the balloon so as to blow up the balloon. According to each embodiment described above, the method for carrying out the equipment from a containment building can shorten the working time for carrying out the RPV from the containment building and reduce the number of casks used for the work. Furthermore, because pipes and structures connected to an RPV are carried out together with the RPV from the containment building, it is possible to shorten the time for removing those pipes and structures. In addition, because the RPV to be carried out is also used as a container for carrying out those pipes and structures, it is possible to reduce the number of containers required to carry out those pipes and structures from the containment building, as well as to reduce the working time for placing those pipes and structures in the container and inspecting its surface contamination. In addition, because the number of containers is reduced, both time and space for transporting storage containers to a storage can be reduced. Consequently, it is possible to reduce the working period for replacing the RPV, as well as to reduce the shut-down period of the object nuclear power plant related to the RPV replacement work. It is also possible to reduce the costs for transporting and storing wastes of pipes and structures after they are cut off and removed. Although the present invention applies to a work for carrying out an RPV in an RPV replacement work in each of the above embodiments, the present invention may also apply to a work for carrying out large equipment (including the RPV) exposed in a nuclear power plant to be disused. Although each of the above embodiments applies to the replacement of a BWR including an RPV, the embodiments may also apply to the replacement of a reactor pressure vessel of a PWR and/or the replacement of a steam generator of the PWR. |
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039502201 | summary | BACKGROUND OF THE INVENTION The present invention relates to recirculating pumps for nuclear reactors in general, and more particularly to improvements in internal primary pumps for recirculation of coolant in pressure vessels of boiling water reactors. It is already known to provide in a nuclear reactor plant a battery of internal recirculating pumps which extend into the bottom wall of the pressure vessel in a boiling water reactor. Such recirculating pumps are installed independently of the feed water supply lines and are driven by normal electric motors or by so-called wet or canned motors known from the art of glandless recirculating pumps. A drawback of such constructions is that the pressure vessel of the reactor must be provided with several sets of openings, namely, those for the introduction of portions of internal recirculating pumps and those for the admission of feed water. Moreover, conventional glandless recirculating pumps constitute expensive auxiliary aggregates especially since, for the reasons of safety, each such pump normally embodies a motor generator with flywheel for supplying additional energy in order to prolong the period of deceleration of the pump rotor in the event of current failure. Gradual deceleration is desirable and necessary in order to insure satisfactory cooling of the pressure vessel. In the absence of aforementioned generators, the rotor of the pump would be arrested practically immediately upon opening of the motor circuit. It was already proposed to employ in a boiling water reactor one or more internal recirculating pumps which are driven by water turbines and are mounted in such a way that feed water enters the pressure vessel by way of openings which are provided for introduction of portions of internal recirculating pumps. Reference may be had to German Offenlegungsschrift No. 1,921,903. This renders it possible to reduce the number of openings in the pressure vessel and the number of conduits for admission of feed water. However, such pressure vessels must be equipped with bypass lines. Another drawback of the just described recirculating systems is that the regulation of water turbines is extremely complex, mainly because the blades of the turbine are not readily accessible for adjustment of their angles. SUMMARY OF THE INVENTION An object of the invention is to provide the pressure vessel of a reactor, especially the pressure vessel of a boiling water reactor, with one or more novel and improved internal primary recirculating pumps which can be installed in openings for the admission of feed water and whose parts are more readily accessible than in heretofore known pressure vessels. Another object of the invention is to provide a novel and improved internal primary recirculating pump for use in the pressure vessel of a boiling water reactor. A further object of the invention is to provide an internal recirculating pump with novel and improved means for prolonging the interval of deceleration upon opening of the circuit of the pump motor. An additional object of the invention is to provide an internal recirculating pump with a novel and improved turbine. The invention is embodied in a nuclear reactor, particularly in a boiling water reactor, which comprises a pressure vessel having one or more openings for admission of feed water and an internal primary recirculating pump for each such opening. Each pump comprises a body a portion of which extends into the respective opening and is formed with an annular chamber for admission of feed water, a shaft which is rotatably mounted in the body and extends into the opening, a rotor which is directly or indirectly carried by the shaft in close proximity to the annular chamber and normally receives torque from the shaft to convey feed water from the annular opening into the pressure vessel, an electric motor mounted in the body and arranged to drive the shaft, and a single-stage or multi-stage turbine, mounted on the shaft, preferably at the suction side of the rotor, to prolong the deceleration of the rotor to zero speed and to thus lengthen the interval of cooling in the event of failure of current supply to the motor. The turbine may serve as a carrier for the pump rotor, and the pressure chamber of the turbine is preferably in communication with the pressure chamber of the rotor by way of an annular flow-restricting bypass which is defined by the pump body and the rotor. The novel features which are considered as characteristic of the invention are set forth in particular in the appended claims. The improved recirculating pump itself, however, both as to its construction and its mode of operation, together with additional features and advantages thereof, will be best understood upon perusal of the following detailed description of certain specific embodiments with reference to the accompanying drawing. |
052316556 | claims | 1. A collimator for collimating radiation beams having a predetermined wavelength distribution emitted from a radiation point source comprising: a plurality of collimator plates stacked together immediately adjacent to one another to form a collimator body, said collimator body being adapted to be situated adjacent to an array of radiation detector elements and having a first surface disposed closest to said array of radiation detector elements; each of said collimator plates comprising radiation absorbent material and having passages therein corresponding to a selected pattern; said plurality of collimator plates being situated so that each of said passages in conjunction with respective passages in adjoining collimator plates form a plurality of respective channels disposed through said collimator body, each of said channels extending from one of said passages opening on said first surface to a respective one of said passages opening on said second surface and having contiguous sidewalls comprising said radiation absorbent material disposed along its length, the passages of each of said channels opening on said second surface being in substantial alignment with a respective one of said detector elements, the longitudinal axis of each of said channels having a selected orientation angle substantially aligned with a direct beam path between said point source and the respective detector element adjoining said channel. a radiation point source; a radiation detector comprising an array of detector elements, said array being disposed to detect radiation emitted from said point source; and a collimator disposed between said detector element array and said radiation point source, said collimator comprising a plurality of collimator plates stacked immediately adjacent to one another, each of which has a selected pattern of passages therein, said plurality of collimator plates being joined together so that each of said passages in conjunction with respective passages in an adjoining collimator plate form a plurality of channels through said collimator to pass radiation emitted by said point source to respective ones of said detector elements, each of said channels having contiguous sidewalls comprising said radiation absorbent material disposed along its length and further having respective longitudinal axes aligned along respective selected orientation angles, said orientation angles corresponding to respective direct paths from said point sour e to respective ones of said detector elements. 2. The collimator of claim 1 wherein each of said passages has substantially similarly-shaped sidewalls. 3. The collimator of claim 1 wherein said radiation absorbent material is selected to substantially absorb radiation of the wavelength distribution emitted by said radiation point source. 4. The collimator of claim 3 wherein said radiation absorbent material comprises one of the group consisting of tungsten, gold, and lead. 5. The collimator of claim 3 wherein each of said collimator plates comprises a sheet of said radiation absorbent material. 6. The collimator of claim 3 wherein said collimator plates each comprise a photosensitive material substrate overlaid with said radiation absorbent material. 7. The collimator of claim 1 wherein the cross-sectional shape of each of said channels corresponds to the cross-sectional shape of said detector element respectively aligned therewith. 8. The collimator of claim 1 wherein said selected pattern of passages in said collimator plates corresponds to the arrangement of said radiation detector elements in said detector array. 9. The collimator of claim 8 wherein said detector elements are arranged in a two-dimensional array. 10. The collimator of claim 1 wherein the respective selected orientation angles of said channels range between about 0.degree. and 10.degree.. 11. A radiation imaging device comprising: 12. The device of claim 11 wherein said radiation point source comprises an x-ray source. 13. The device of claim 12 wherein said collimator plates are comprised of an x-ray absorbent material. 14. The device of claim 13 wherein said radiation absorbent material comprises a material chosen from the group consisting of tungsten, gold and lead. 15. The device of claim 13 wherein each of said collimator plates comprises a photosensitive glass substrate overlaid with said radiation absorbent material. 16. The device of claim 13 wherein each of said collimator plates comprises a sheet of said radiation absorbent material. 17. The device of claim 11 wherein the respective selected orientation angles of said channels range between about 0.degree. and 10.degree.. 18. The device of claim 11 wherein said detector elements are arranged in a two-dimensional array. |
summary | ||
049960172 | claims | 1. A neutron generator tube comprising: (a) a hermetically sealed housing containing an ionizable gas; (b) an axial recess in one end of said housing adapted to receive a magnet; (c) a ring anode axially oriented within said housing adjacent to said recessed end for accepting said magnet; (d) an axially oriented thermal conductor cathode penetrating through the other end of said housing wherein the inner surface of said cathode contains a target; (e) an ion screen ensemble near said anode ring and between said anode ring and cathode target wherein said ion screen ensemble contains an axially positioned gridded aperture wherein said aperture is substantially smaller than said anode ring and target and wherein said ion screen ensemble further comprises: (f) an electron shield near said cathode target and between said ion screen ensemble and cathode target wherein said electron shield contains an axially positioned aperture. (a) providing said neutron generator tube with an axially aligned recess adapted to accept a removable magnet wherein said recess terminates internal to said tube substantially adjacent to one side of said ring anode; and (b) a removable magnet adapted to fit into said recess. 2. A neutron generator tube of claim 1 further comprising a removable magnet adapted to fit into said axial recess. 3. A neutron generator tube of claim 2 wherein said removable magnet is a samarium/cobalt magnet. 4. A neutron generator tube of claims 1 or 2 wherein said thermal conductor cathode has a cross-section of substantially the same size as the target and wherein the end of said cathode containing the target is within said housing and the other end extends outside said housing such as to remove thermal energy from said target. 5. A neutron generator tube of claim 4 wherein said thermal conductor cathode is a OFHC copper rod having a film of titanium deposited on a polished end of said rod to serve as said target. 6. A neutron generator tube of claims 1 or 2 wherein said ion screen ensemble is an electrically conductive, grounded surface. 7. A neutron generator tube of claim 6 wherein said gridded aperture is a screen of etched tungsten of about 0.002 to 0.005 inch thick with the openings representing about 90 percent of the surface area. 8. In a neutron generator tube containing an ion source involving a ring anode and cathode sealed in an ionizable gas wherein a magnetic field axially aligned with said ring anode is used, the improvement comprising: 9. A neutron generator tube of claim 8 wherein said removable magnet is a samarium/cobalt magnet. 10. In a neutron generator tube containing an ion source involving a ring anode and a cathode target sealed in an ionizable gas, the improvement comprising: a thermally conductive target support having a cross-section of substantially the same size as the target and wherein the end of said support containing the target is within said tube and the other end extends outside the tube such as to remove thermal energy from said target, said thermally conductive target support further comprising an OFHC copper rod having a film of titanium deposited on the polished end of said rod to serve as said target. 11. In a neutron generator tube containing an ion source involving a ring anode with a magnetic field axially aligned with said ring anode and a cathode target axially aligned to said ring anode sealed in an ionizable gas, the improvement comprising: an ion screen ensemble positioned between said ring anode and said cathode target containing an axially aligned gridded aperture wherein said ion screen ensemble is a multilayered structure consisting of a highly polished outer layer facing said target, a ferromagnetic structural inner layer, and a metallic gridded intermediate layer covering said aperture. 12. A neutron generator tube of claim 11 wherein said metallic grid is a screen of etched tungsten of about 0.002 to 0.005 inch thick with the openings representing about 90 percent of the surface area. |
summary | ||
abstract | A particle beam emitter has a hollow particle beam tube having a first end portion, a second end portion, and a longitudinal axis. An electromagnetic system that includes a voltage supply is electrically coupled to the hollow particle beam tube and is configured to generate a primary electrical current flowing axially in the hollow particle beam tube from the first end portion towards the second end portion. A primary magnetic field associated with the primary electrical current is operable to induce a secondary electrical current in a plasma located within the hollow particle beam tube, the secondary electrical current flowing generally axially within the plasma and causing the plasma to contract inwardly towards the longitudinal axis. |
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claims | 1. A lifting support for a boiling water reactor nuclear fuel assembly comprising:a grappling head configured to allow attachment of a lifting device;a body with an upper end and a lower end, the upper end connected to the grappling head, the body configured to be inserted into a water channel of the boiling water reactor nuclear fuel assembly; andan end connected to the lower end, the end configured to be accepted by the water channel of the nuclear fuel assembly, wherein the body defines an interior volume and the lifting support further comprises:an inner rod with a first end and a second end, the inner rod positioned in the interior volume;an actuator connected to the inner rod at the first end, the actuator configured to actuate and move the inner rod; anda ball lock end connected to the second end, the ball lock end configured to be actuated by the inner rod, the end further configured to be inserted into a water channel of a boiling water nuclear fuel assembly in a non-extended position and provide a load path for lifting the fuel assembly when in an extended position, the ball lock end positioned at the end connected to the lower end, wherein the lifting support has a length sufficient to insert the ball lock end to an elevation of a single lower tie plate of the fuel assembly. 2. The lifting support according to claim 1, wherein the lifting support is made from stainless steel. 3. The lifting support according to claim 2, wherein the stainless steel is type 18-8. 4. The lifting support according to claim 1, further comprising:a spacer located on the body, the spacer configured to limit direct contact between the body of the lifting support and the fuel assembly. 5. The lifting support according to claim 4, wherein the spacer is plastic. 6. The lifting support according to claim 4, wherein the spacer is removable from the body of the lifting support. 7. The lifting support according to claim 1, wherein the ball lock end is configured to engage the water channel in the extended position through a sheer connection. 8. The lifting support according to claim 1, wherein the actuator is a remote controlled actuator. 9. A fuel assembly for a boiling water reactor comprising:a fuel channel configured to define a volume, the channel having a lower end and an upper end, the lower and the upper ends open, the fuel channel further configured to channel a flow of coolant;a plurality of fuel rods each with a lower end and an upper end placed within the volume, the plurality of fuel rods containing fissile material;a lower single tie plate configured to support the plurality of rods, the lower tie plate configured to abut an interior side of the fuel channel lower end and allow flow of the coolant through the single lower tie plate through the channel;a removable upper tie plate configured to be attached to the fuel channel, the removable upper tie plate configured to restrict movement of the plurality of fuel rods;a water channel placed in the volume and connected to the single lower tie plate, the water channel configured to channel coolant flow from the single lower tie plate, the water channel further configured with an attachment end;a removable lifting support configured with a grappling head to allow attachment to a lifting device, a body with an upper end and a lower end, the upper end connected to the grappling head, the body configured to be inserted into the water channel of the boiling water reactor nuclear fuel assembly, and an end connected to the lower end, the end configured to be accepted and retained by the water channel attachment end, the end further configured to be disengaged from the water channel, wherein the body has an interior volume and the lifting support further comprises:an inner rod with a first end and a second end, the inner rod positioned in the interior volume;an actuator connected to the inner rod at the first end, the actuator configured to actuate and move the inner rod; anda ball lock end connected to the second end, the ball lock end configured to be actuated by the inner rod, the end further configured to be inserted into a water channel of a boiling water nuclear fuel assembly in a non-extended position and provide a lead path for lifting the fuel assembly when in an extended position; the ball lock end positioned at the end connected to the lower end, wherein the lifting support has a length sufficient to insert the ball lock end to an elevation of the single lower tie plate of the fuel assembly;a plurality of spacers positioned between the single lower tie plate and the removable upper tie plate, the plurality of spacers configured to maintain individual fuel rod positions of the plurality of rods and channel coolant flow; anda lower nozzle configured to channel fluid to the single lower tie plate and the water channel, the nozzle connected to the fuel channel. 10. The fuel assembly according to claim 9, wherein the lifting support is made from stainless steel. 11. The fuel assembly according to claim 9, wherein the fuel channel and the removable upper tie plate have matching slots and further comprising:a bolt configured with a head and an end, wherein the bolt is positioned through the slot in the fuel channel and the removable upper tie plate to provide a connection between the fuel channel and the removable upper tie plate. 12. The fuel assembly according to claim 11, further comprising:a screw configured to be attached to the bolt, wherein the bolt is configured with a head that has a larger diameter than the slot in the fuel channel, the bolt further configured with a hole to accept the screw to form a connection;a top plate configured with a hole wherein the screw is inserted, the top plate configured to be removably attached to the fuel channel upper end through an attachment; anda member attached to the fuel channel, the member configured to be attached to a load lifting device. |
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claims | 1. A process for recovering uranium and molybdenum-99 (Mo-99) from an irradiated solid target, comprising:irradiating a solid target comprising uranium to produce the irradiated solid target having fission products comprising Mo-99;dissolving the irradiated solid target to form a first solution and conditioning the first solution to comprise a first nitric acid concentration and a first uranium concentration;oxidizing the Mo-99 by adding an inorganic oxidant to the first solution to provide a second solution in which the Mo-99 is in a +VI oxidation state;contacting the second solution with a solid sorbent, whereby uranium remains in the second solution while the Mo-99 is bound to the solid sorbent;adjusting the first nitric acid concentration and the first uranium concentration in the second solution to provide a second nitric acid concentration and a second uranium concentration suitable for formation of uranyl nitrate hydrate crystals;inducing formation of uranyl nitrate hydrate crystals; andseparating the uranyl nitrate hydrate crystals from the second solution. 2. The process of claim 1, wherein the inducing formation of uranyl nitrate hydrate crystals comprises cooling the second solution to a temperature effective for the formation of the uranyl nitrate hydrate crystals. 3. The process of claim 1, wherein the inducing formation of uranyl nitrate hydrate crystals comprises evaporating the second solution under reduced pressure. 4. The process of claim 1, further comprising purifying the uranyl nitrate hydrate crystals after separating the uranyl nitrate hydrate crystals. 5. The process of claim 1, wherein the uranium comprises low-enriched uranium (LEU). 6. A process for recovering uranium and Mo-99 from an irradiated solid target, comprising:irradiating a solid target comprising uranium to produce the irradiated solid target having fission products comprising Mo-99;dissolving the irradiated solid target to form a first solution and conditioning the first solution to comprise a nitric acid concentration of about 0.01 M to about 2 M, and a uranium concentration of about 50 gU/L to about 350 gU/L;oxidizing the Mo-99 by adding an inorganic oxidant to the first solution to provide a second solution in which the Mo-99 is in a +VI oxidation state;contacting the second solution with a solid sorbent, whereby uranium remains in the second solution while the Mo-99 is bound to the solid sorbent;adjusting the nitric acid concentration in the second solution to about 4 M to about 8 M, and the uranium concentration in the second solution to about 350 gU/L to about 650 gU/L;inducing formation of uranyl nitrate hydrate crystals; andseparating the uranyl nitrate hydrate crystals from the second solution. 7. The process of claim 6, wherein the inducing formation of uranyl nitrate hydrate crystals comprises cooling the second solution to a temperature effective for the formation of the uranyl nitrate hydrate crystals. 8. The process of claim 6, wherein the inducing formation of uranyl nitrate hydrate crystals comprises evaporating the second solution under reduced pressure. 9. The process of claim 6, further comprising purifying the uranyl nitrate hydrate crystals after separating the uranyl nitrate hydrate crystals. 10. The process of claim 6, wherein the uranium comprises LEU. 11. The process of claim 1, further comprising recycling the nitric acid. 12. The process of claim 6, further comprising recycling the nitric acid. |
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053234359 | claims | 1. A support for a control rod drive housing in a boiling water reactor comprising: a first means for supporting a control rod drive; and a second means for supporting said control rod drive, said second means disposed on said first supporting means wherein said second means is movable on said first supporting means from a support position to a non-support position, whereby said control rod drive is supported by said first and second supporting means when said second means is in said support position and said control rod drive is not supported by said first and second supporting means and can be replaced without removing said second supporting means from said first support means when said second supporting means is in said non-support position. a plurality of first support members provided in rows on opposing sides of a lower portion of a plurality of control rod drives; and a plurality of second support members disposed on said first support member, said second support member having a plurality of extension members for supporting said control rod drives, wherein said second support member is movable from a support position to a non-support position, whereby said control rod drive is supported by said extension members when said second means is in said support position and said control rod drive is not supported by said extension members and can be replaced without removing said second support means from said first support means when said second support means is in said non-support position. 2. A control rod drive housing support according to claim 1, wherein said first supporting means comprises a plurality of support members provided in rows on opposing sides of a lower portion of a plurality of control rod drives. 3. A control rod drive housing support according to claim 2, wherein said second supporting means comprises a plurality of second support members disposed on said first support members, each of said second support members having a plurality of extension members for supporting said control rod drives, whereby said control rod drive is supported by said extension members when said second support is in said support position and said control rod drive is not supported by said extensions members when said second support is in said non-support position. 4. A control rod drive housing support according to claim 3, wherein said second support is rotatably disposed on said first support member, whereby said extensions are moveable rotatably from said support position to said non-support position and from said non-support position to said support position. 5. A control rod drive housing support according to claim 4, wherein said of first support members have a plurality of hubs for receiving said plurality of second support members rotatably thereon. 6. A control rod drive housing support according to claim 5, wherein said second support member has a hole therethrough for receiving one of said plurality of hubs rotatably therein. 7. A control rod drive housing support according to claim 6, wherein said plurality of extension members on each of said second support member is four, each of said extension members supporting a quadrant of said control rod drive, whereby one extension from four second support members support one control rod drive. 8. A control rod drive housing support according to claim 3, wherein said first support member has a first hole therethrough and said second support member has a second hole therethrough for receiving an instrument to be supported in said control rod drive housing, said second support member having a rim extending beyond said second hole which is rotatably mounted in said first hole of said first support member, whereby said first support member supports both the second support member an said instrument. 9. A control rod drive housing support according to claim 8, wherein said plurality of extension members on each of said second support member is four, each of said extension members supporting a quadrant of said control rod drive, whereby one extension member from four second support members support one control rod drive. 10. A control rod drive housing support according to claim 1, wherein said plurality of extension members on said second support member are removable, whereby said extensions are disposed on said first support member when in said support position and are removed from said first support member when in said non-support position. 11. A control rod drive housing support according to claim 10, wherein said plurality of extension members on said second support member is two, each of said extension members supporting half of said control rod drive. 12. A support for a control rod drive housing in a boiling water reactor comprising: 13. A control rod drive housing support according to claim 12, wherein said second support is rotatably disposed on said first support member, whereby said extensions are moveable rotatably from said support position to said non-support position and from said non-support position to said support position. 14. A control rod drive housing support according to claim 13, wherein said of first support members has a plurality of hubs for receiving a plurality of second support members rotatably thereon. 15. A control rod drive housing support according to claim 14, wherein said second support member has a hole therethrough for receiving one of said plurality of hubs rotatably thereon. 16. A control rod drive housing support according to claim 14, wherein said plurality of extension members on said second support member is four, each of said extension members supporting a quadrant of said control rod drive, whereby one extension from four second support members support one control rod drive. 17. A control rod drive housing support according to claim 13, wherein said first support member has a first hole therethrough and said second support member has a second hole therethrough for receiving an instrument to be supported in said control rod drive housing, said second support member having a rim extending beyond said second hole which is rotatably mounted in said first hole of said first support member, whereby said first support member supports both the second support member an said instrument. 18. A control rod drive housing support according to claim 17, wherein said plurality of extension members on said second support member is four, each of said extension members supporting a quadrant of said control rod drive, whereby one extension from four second support members support one control rod drive. 19. A control rod drive housing support according to claim 12, wherein said plurality of extension members on said second support member are removable, whereby said extensions are disposed on said first support member when in said support position and are removed from said first support member when in said non-support position. 20. A control rod drive housing support according to claim 19, wherein said plurality of extension members on said second support member is two, each of said extension members supporting half of said control rod drive. |
048428094 | abstract | A rod arraying system for nuclear fuel rods or burnable poison rods. A support structure positions a consolidation canister vertically to receive the rods. A rod loading pattern provides a gap between rods to allow projections to hold each rod in position. One or more rod arraying devices positioned vertically along and inclined toward the canister provide a triangular pitch loading configuration. A guide assembly slidably mounted on the mounting base of the rod arraying device has two guide plates with scalloped edges. The guide plates are offset relative to each other with one being cycled back and forth by an air cylinder to allow individual loading of fuel rods one row at a time. The inclined weight of the guide plates keeps them in position against the rods to prevent unwanted shifting of rods during the loading operation. |
046882421 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT As shown in the block diagram of FIG. 1, an X-ray imaging system according to an embodiment of the present invention comprises X-ray tube 1, X-ray generation unit 2, collimator 3, X-ray mask member 4, drive unit 5, X-ray image detector 6, camera 7, A/D (analog-to-digital) converter 8, memories 9 to 14, calculating sections 15 and 16, calculating/controlling unit 17, D/A (digital-to-analog) converter 18, display 19, and data bus 20. When X-ray tube 1 is driven by X-ray generation unit 2, X-rays are emitted from tube 1 toward object (e.g., a patient) P. A radiation field of X-rays emitted from tube 1 is restricted by collimator 3. The X-rays, the radiation field of which is defined by collimator 3, become incident on object P through X-ray mask member 4, which partially shields X-rays. Mask member 4 comprises X-ray transmitting plate 31 of, e.g., a thin acrylic resin plate having good X-ray transmittance, on which a plurality of small segments (e.g., lead segments) 32 of an X-ray shielding material are attached at equal intervals, as shown in FIG. 2. Each lead segments 32 has a size of 2 mm.times.2 mm, for example. Mask member 4 locally shields the X-rays with a predetermined pattern constituted by the plurality of lead segments 32. Mask member 4 is moved by drive unit 5 to be inserted in or removed from the X-ray radiation field. In addition, when member 4 is located in the X-ray radiation field, it can be displaced to different positions therein upon operation of drive unit 5. X-ray image detector 6 comprises, for example, an image intensifier (I.I.), and detects X-ray image data transmitted through object P to convert it into a visible image. The visible X-ray image obtained by detector 6 is converted into an electrical signal by camera (e.g., TV camera) 7, and is then converted into a digital signal by A/D converter 8. 1st memory 9 stores transmission X-ray data (original image data XO) transmitted through object P when mask member 4 is located outside the X-ray radiation field. 2nd memory 10 stores X-ray image data (first masked X-ray data MA) obtained when mask member 4 is located at a first predetermined position in the X-ray radiation field. 3rd memory 11 stores X-ray data (second masked X-ray data MB) obtained when mask member 4 is located at a second predetermined position in the X-ray radiation field separated from the first predetermined position. 1st calculating section 15 performs subtraction processing of original image data XO stored in 1st memory 9 and first masked X-ray data MA stored in 2nd memory 10 for each corresponding pair of pixels, and stores resultant first subtraction data SA in 4th memory 12. Calculating section 15 also performs subtraction processing of original image data XO and second masked X-ray data MB stored in 3rd memory 11 for each corresponding pair of pixels, and stores resultant second subtraction data SB also in 4th memory 12. Calculating section 15 also performs subtraction processing of original image data XO and scattered X-ray distribution data DS (to be described later) indicating a scattered X-ray intensity distribution for each corresponding pair of pixels, and outputs the processing result to D/A converter 18. Calculating/controlling unit 17 comprising, e.g., a CPU (central processing unit), controls the operation of the entire system (e.g., the operation of X-ray generation unit 2 for driving X-ray tube 1, drive unit 5 for driving mask member 4, and camera 7, read/write control of all the memories, and the like). Unit 17 calculates a central address for each region shielded with lead segments 32 of mask member 4 (i.e., an X-ray shielding region) in a memory, displacement of the address due to movement of mask member 4, and an average value of X-ray intensity data in the X-ray shielding regions, i.e., scattered X-ray component data. 5th and 6th memories 13 and 14 store transient calculation data from unit 17. 2nd calculating section 16 performs data interpolation using, e.g., a SYNC function, on the basis of data associated with the X-ray shielding regions calculated by unit 17, thus obtaining scattered X-ray distribution data. Various data communication between respective memories and calculating sections is made through data bus 20. The operation of the system with the above arrangement will now be described. At the beginning of an imaging operation, mask member 4 is located outside the X-ray radiation field. In this state, unit 17 causes X-ray generation unit 2 to radiate X-rays from X-ray tube 1 toward object P. In synchronism with X-ray radiation, unit 17 controls camera 7 to obtain original image data XO through X-ray image detector 6. Acquired original image data XO is converted into digital data by A/D converter 8, and is written in 1st memory 9. Next, drive unit 5 is driven under the control of unit 17, so that mask member 4 is moved to a first position in the X-ray radiation field. In this state, X-rays are emitted from X-ray tube 1, and first masked X-ray data MA is acquired by camera 7 through detector 6. Data MA is then written in 2nd memory 10 through A/D converter 8. The position of mask member 4 is shifted in the same plane to a second position in the X-ray radiation field slightly shifted from the first position. Upon re-radiation of X-rays in this state, second masked X-ray data MB is acquired by camera 7. Data MB is then written in 3rd memory 11 through A/D converter 8. The relationship between X-ray data MA and MB, written in 2nd and 3rd memories 10 and 11, and mask member 4 will now be described in detail with reference to FIGS. 3 to 6. FIG. 3 is an illustration for explaining acquisition of masked X-ray data, and FIG. 4 is a graph showing an X-ray intensity distribution detected when X-rays are radiated in the state shown in FIG. 3. When X-rays XR are radiated in the state wherein mask member 4 is positioned in the X-ray radiation field, an X-ray intensity distribution of masked X-ray data along line A--A' detected by detector 7 is as shown in FIG. 4. The X-ray intensity distribution in FIG. 4 shows steep dips at positions indicated by a', b', c', d', and e'. These dips represent that X-ray XR is locally shielded by lead segments 32 (i.e., segments 32a, 32b, 32c, 32d, and 32e) along line A--A'. Although X-ray XR is shielded by lead segments 32a, 32b, 32c, 32d, and 32e, values at distal ends of the dips are not zero because scattered X-rays are detected. Thus, level and distribution of the scattered X-ray components can be calculated from the values of the dips. In this calculation, as the number of dips, i.e., the number of lead segments 32 provided in mask member 4, increases, detection precision of the scattered X-ray components can be improved. However, if the number of dips, i.e., the number of lead segments 32, is too large, the amount of X-rays shielded by segments 32 increases, and the amount of X-rays transmitted through object P and landing on detector 6 therefore decreases, resulting in a smaller amount of scattered X-ray component data than actually exists. In theory, as the number of X-ray shielding regions included in an imaging region increases, more effective scattered X-ray component data could be obtained. However, in practice, a distribution density of the X-ray shielding regions must be set below a given level. Alternatively, the size of segments 32 can be reduced. However, in order to allow detection of the dips, segments 32 cannot be smaller than a given size. Because, as seen from FIG. 8B, in general, the waveform of the X-ray intensity signal is distorted and stretched. It is therefore necessary to distinguish the desirable signal belonging to the X-ray shielded portion (i.e., the scattered X-ray signal) from the signal belonging to the other portion (i.e., the primary X-ray signal and a part of the scattered X-ray signal). Therefore, as seen FIG. 8A to 8C, the X-ray intensity data of the portion which is not shielded by the lead piece 32 has a higher level than the threshold level and is converted into the digital "0" level signal. The X-ray intensity data of the portion which is shielded by the lead piece 32 has a lower level than the threshold level and is converted into the digital "1" level signal. This bilevel quantization is carried out in 1st calculating section 15. In the system of this embodiment, instead of increasing the number of segments 32, X-ray mask member 4 is displaced to, e.g., two different positions in the X-ray radiation field, and X-ray data MA and MB are obtained during respective X-ray radiations at the two positions, thereby providing the same effect as when the number of segments 32 is increased. FIGS. 5A to 5C and FIG. 6 are illustrations for explaining movement of mask member 4 in the X-ray radiation field in the system of this embodiment. If mask member 4 in a state shown in FIG. 5A is parallel-moved by distance dx in the X direction, a state shown in FIG. 5B is obtained. Assuming that the coordinates of lead segment 32p in FIG. 5A are (x, y), those in FIG. 5B are (x+dx, y). If mask member 4 in the state in FIG. 5A is parallel-moved by distance dy in the Y direction, a state shown in FIG. 5C is obtained, and the coordinates of segment 32p are (x, y+dy). X- and Y-movements can be freely combined. FIG. 6 illustrates the position of member 4 before and after it is moved in the X and Y directions by distances dx and dy, respectively, when these movements overlap each other. Referring to FIG. 6, after the two parallel-movements, the coordinates (x, y) of segment 32p before the movements are changed to be (x+dx, y+dy). When member 4 is moved by distance DX in the X direction, the coordinates of segment 32p are the same as they were before the movements. Therefore, the amount of movement in the X direction can be smaller than distance DX and that in the Y direction can be smaller than distance DY, so that the coordinates of segment 32p are the same as those of lead segment 32r before the movements. Within this limited range, segment 32p can be moved to any location encompassed by segments 32p, 32q, 32r, and 32s, and the object of this embodiment can be satisfactorily achieved thereby. Masked X-ray data MA and MB acquired by displacing mask member 4 to different positions as above are subjected to subtraction with original image data XO. 1st calculating section 15 executes subtraction processing of original image data XO stored in memory 9 and X-ray data MA stored in memory 10 for each corresponding pair of pixels. The subtraction result (i.e., subtraction data SA) from section 15 is then stored in memory 12. Calculating/controlling unit 17 calculates central addresses of the X-ray shielding regions in X-ray data MA, and scattered X-ray component data corresponding to the respective X-ray shielding regions in accordance with subtraction data SA stored in memory 12. The data is then written in memory 13. This data will be referred to as X-ray shielding region data AA hereinafter. Calculating section 15 also executes subtraction processing of original image data XO and X-ray data MB stored in memory 11. The subtraction result (i.e., subtraction data SB) from section 15 is also stored in memory 12. Unit 17 calculates central addresses of the respective X-ray shielding regions in data MB, and scattered X-ray component data corresponding to the respective X-ray shielding regions in accordance with subtraction data SB stored in memory 12. The data from unit 17 is then stored in memory 14. This data will be referred to as X-ray shielding region data AB hereinafter. Unit 17 calculates a distance between the central addresses of the X-ray shielding regions (i.e., displacement of member 4) based on data AA and AB respectively stored in memories 13 and 14, and outputs the result to 2nd calculating section 16 together with data AA and AB. Calculating unit 16 executes data interpolation using a SINC function based on the output data from unit 17, thus obtaining scattered X-ray distribution data indicating the scattered X-ray intensity distribution. The scattered X-ray distribution data is then supplied to section 15 to be subjected to subtraction processing with original image data XO stored in memory 9. After the subtraction processing, the scattered X-ray components contained in original image data XO can be eliminated, and image data corresponding to an image formed only by direct X-ray components can be obtained. The image data is then supplied to display 19 through D/A converter 18, to be displayed as an image. Since the image displayed on display 19 is free from the influence of scattered X-rays, it has high contrast and sharpness. Note that the scattered X-ray distribution data is subtracted from original X-ray data XO stored in memory 9 by section 15. Alternatively, after acquiring the masked X-ray data, X-ray radiation can be performed while mask member 4 is located outside the X-ray radiation field, so as to obtain transmission X-ray data. The transmission X-ray data is used instead of original image data XO, and the above subtraction can be performed. Once the scattered X-ray distribution data is obtained, it can be used to eliminate the scattered X-ray components from transmission X-ray data during the subsequent X-ray radiation as long as the position or size of the X-ray radiation field or X-ray radiation conditions remain substantially the same. FIGS. 7A to 7C respectively show timings of X-ray radiation, image detection of detector 6, and movement of mask member 4. In this embodiment, camera 7 comprises a vidicon TV camera having an accumulation type target. X-ray tube 1 radiates X-ray pulses at equal time intervals, as shown in FIG. 7A. A transmission X-ray image accumulated on the target of camera 7 through detector 6 during the X-ray radiation is read out during an interval between two X-ray radiations, as shown in FIG. 7B. Mask member 4 is moved during such an interval, i.e., during the readout interval of camera 7. In this case, when member 4 is located outside the radiation field, a first X-ray radiation is performed to acquire original image data XO. During the readout interval of data XO from camera 7, mask member 4 is moved to a first position in the radiation field. After this movement, a second X-ray radiation is performed to acquire masked X-ray data MA. During the readout interval of data MA from camera 7, mask member 4 is moved to a second position in the radiation field. After completion of the movement, a third X-ray radiation is performed to acquire masked X-ray data MB. During the readout interval of data MB from camera 7, mask member 4 is moved outside the radiation field. The scattered X-ray distribution data is calculated after the above sequence. Transmission X-ray data obtained from repetitive X-ray radiation can be corrected with the same scattered X-ray distribution data unless a position or direction of the object, a position or size of the radiation field, and X-ray radiation conditions are greatly changed. Masked X-ray data MA and MB need only provide positions and values of dips caused by the X-ray shielding regions, and the number of X-ray shielding regions of member 4 can be reduced because member 4 itself is displaced. Even if data MA and MB, subtraction data SA and SB, and X-ray shielding region data AA and AB corresponding thereto express 1-frame image data in a smaller (rougher) matrix size than that of original image data, detection precision will substantially not be degraded. Therefore, the capacities, per frame of an image, of memories 10 to 14 for storing data MA, MB, SA, SB, AA, and AB can be smaller than that of memory 9. If original image data is constituted by 512.times.512 pixels, data MA, MB, SA, SB, AA, and AB can often be expressed by 128.times.128 or 256.times.256 pixels. In this case, the correspondence between pixels of the respective image data must be corrected during the above calculations. With the system of this embodiment, scattered X-ray distribution data is obtained based on X-ray shielding data acquired when mask member 4 is located in the X-ray radiation field, and scattered X-ray components contained in the original image data are eliminated through subtraction processing of the scattered X-ray distribution data and the original image data, thus displaying an image of high contrast and sharpness. In particular, the system can provide the same effect as when the number of lead segments 32 is increased, since a plurality of masked X-ray data acquired when mask member 4 is at different positions in the X-ray radiation field are combined. Therefore, measurement precision of the scattered X-ray component distribution can be improved without increasing the number of segments 32. The present invention has been described in connection with a particular embodiment. However, the present invention is not limited to the above embodiment, and various changes and modifications may be made within the spirit and scope of the invention. For example, in the above embodiment, movement of mask member 4 (i.e., displacement thereof into, from, or within the X-ray radiation field) is performed through drive unit 5 under the control of unit 17. However, mask member 4 can be moved manually. In the system of the above embodiment, original image data is acquired when mask member 4 is out of the X-ray radiation field, and X-ray shielding region data is then acquired when it is inside the field. However, the X-ray shielding region data can be obtained first, and thereafter, the original image data can be obtained. |
description | The present disclosure relates to an ion implantation method and apparatus for implanting ions into a substrate using both a ribbon-like (this is called also a sheet-like or a strip-like) ion beam in which, with or without performing an X-direction sweep, a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, and a mechanical scan of the substrate in a direction intersecting with the principal face of the ion beam. In the specification, in order to be easily distinguished from a sweep of an ion beam, an operation of mechanically sweeping a substrate is referred to as a scan. FIG. 9 shows a related-art example of an ion implantation apparatus of this kind. The ion implantation apparatus has a configuration which implants ions into (for example, a whole face of) a substrate (for example, a semiconductor substrate) 2 using both a ribbon-like ion beam 4 and an operation of mechanically scanning the substrate 2 in a direction intersecting with the principal face 4b (see FIG. 11) of the ion beam 4, by a substrate driving device 10. Referring to FIG. 11, for example, the ion beam 4 undergoes a sweep process in the X direction (for example, a horizontal direction) which is based on an electric or magnetic field produced by a beam sweeper (not shown), and has a ribbon-like section shape in which the dimension in the X direction is larger than that in the Y direction (for example, a vertical direction) that is orthogonal to the X direction. For example, the ion beam 4 before the sweep operation has a section shape such as a small oval or circle as indicated by the reference numeral 4a in FIG. 11. Alternatively, without undergoing such sweep a process in the X direction, the ion beam 4 (for example, the ion beam itself derived from an ion source) may have a ribbon-like section shape in which the dimension in the X direction is larger than that in the Y direction. In this example, the substrate driving device 10 has: a holder 12 which holds the substrate 2; a motor 14 which rotates the holder 12 together with the substrate 2 about a center portion 2a of the substrate 2 as indicated by the arrow A (or in the opposite direction) (this motor is referred to as the twist motor in order to be distinguished from a motor 16 which will be described later); and the motor 16 which drives (reciprocally swings) the holder 12 together with the substrate 2 and the twist motor 14 as indicated by the arrow B to change the inclination angle θ of the holder 12 and the substrate 2 (this motor is referred to as the tilt motor in order to be distinguished from the twist motor 14). For example, the inclination angle θ can be changed in a range from 0 deg. (i.e., the state where the holder 12 is vertical) to the vertical to 90 deg. (i.e., the state where the holder 12 is horizontal). The substrate driving device 10 further has a scanning device 18 which mechanically scans the holder 12, the substrate 2, and the like so as to reciprocate between one end (for example, the lower end) 20 of the scan and the other end (for example, the upper end) 22 as indicated by the arrow C, thereby mechanically scanning the substrate 2 in a direction (for example, the Y direction) intersecting with the principal face 4b of the ion beam 4. The scan direction of the substrate 2 is not restricted to the direction of the arrow C (the Y direction). In some cases, the scan may be performed in parallel with the surface of the substrate 2. In the specification, one scan of the substrate 2 means a one-way scan. A substrate driving device having a configuration which is substantially identical with that of the substrate driving device 10 is disclosed in Patent Reference 1. As shown in FIG. 10, for example, replacement of the substrate 2 with respect to the holder 12 (for example, that of an ion-implanted substrate 2 with a substrate 2 before the ion implantation) is performed while the holder 12 is set to a substantially horizontal state at the one end 20 of the scan. In the ion implantation into the substrate 2, in accordance with Expression 1 or an expression which is mathematically equivalent thereto, for example, the scan number of the substrate 2 is calculated by using the beam current of the ion beam 4, the dose amount to the substrate 2, and a reference scan speed which is used as a reference for calculating the scan number of the substrate 2. Usually, the calculated scan number is a mixed decimal with number of digits after the decimal point. Therefore, a scan number in which the digits after the decimal point are truncated, or which is an integer is calculated, and the calculated number is set as a scan number which is practically used. In the case where the calculated scan number is 3.472, for example, 3 is set as the scan number which is practically used. In accordance with Expression 2 or an expression which is mathematically equivalent thereto, for example, the scan speed which is practically used is calculated by the scan number which is practically used. In related-art, the ion implantation is performed on the substrate 2 in accordance with the scan number and scan speed which are calculated in this manner. scan number [ time ] = dose amount [ ions / cm 2 ] × reference scan speed [ cm / sec ] × elementary electric charge [ C ] × coefficient beam current × 10 - 6 [ C / sec ] [ Expression 1 ] scan speed [ cm / sec ] = beam current × 10 - 6 [ C / sec ] × scan number dose amount [ ions / cm 2 ] × elementary electric charge [ C ] × coefficient [ Expression 2 ] In Expressions 1 and 2 described above, the elementary electric charge is 1.602×10−19 [C], and the coefficient is a coefficient which is specific to the ion implantation apparatus. This is applicable also to Expression 4 which will be described later. FIG. 12 shows an example of the scan number and scan speed of the substrate 2 in the case where the related-art ion implantation method (apparatus) is employed. FIG. 12 is a graph showing transitions of the scan number and the scan speed in the case where the dose amount is fixed and the beam current is reduced. The reference scan speed was 320 mm/sec. In order to make the dose amount constant, the scan number is increased in accordance with reduction of the beam current. In the case where the beam current is largely reduced, the scan number is largely increased. Also in the case where the beam current is fixed and the dose amount is increased, a similar tendency is obtained. [Patent Reference 1] JP-A-2004-95434 (Paragraphs [0010] to [0017], FIG. 6) In the vicinities of the ends of the scan of the substrate 2, the substrate must be decelerated and accelerated in order to perform the scan return operation, and a time loss is caused by the deceleration and the acceleration. This will be described in detail with reference to FIG. 13. This time loss is a total of wasted times other than the time of implantation into the substrate 2 (this time occurs twice per scan) during the time for one scan. Most of the time loss consists of the deceleration time before the scan return and the acceleration time after the scan return. The time loss is the sum of these deceleration and acceleration times and an overscan time (this time also occurs twice per scan) for the overscan of the substrate 2 (the operation in which the substrate 2 is overscanned in slightly excess so that the substrate 2 is surely located outside the ion beam 4). For example, the time loss is about 0.4 sec. per scan in the case where the reference scan speed is 320 mm/sec., and about 0.5 sec. per scan in the case where the reference scan speed is 200 mm/sec. In the specification, the term “scan speed” means the scan speed in the implantation time and the overscan time. When the scan number is increased, also the number of decelerations and accelerations of the substrate in the vicinities of the scan ends is increased, with the result that many time losses are accumulated, and the throughput is reduced. Exemplary embodiments of the present invention provide an ion implantation method and apparatus in which accumulation of time losses which mainly consist of deceleration and acceleration times in the vicinities of scan ends is suppressed, so that the throughput is improved. An ion implantation method according to a first aspect of the invention is characterized in that the method includes: a scan speed calculating step of setting an initial value of a scan number of the substrate to 1, and calculating a scan speed of the substrate by using a beam current of the ion beam, a dose amount to the substrate and the initial value of a scan number of the substrate; a scan speed determining step of: determining whether the scan speed of the substrate is within a predetermined allowable scan speed range or not; if the scan speed is within the allowable scan speed range, setting the current scan number and the current scan speed as a practical scan number and a practical scan speed, respectively; if the scan speed is higher than an upper limit of the allowable scan speed range, aborting a process of obtaining the practical scan number and the practical scan speed; and, if the scan speed is lower than a lower limit of the allowable scan speed range, incrementing the scan number by one to calculate a corrected scan number; a corrected-scan speed calculating step of, when the corrected scan number is calculated, calculating a corrected scan speed by using the corrected scan number, the beam current, and the dose amount; a repeating step of, when the corrected scan speed is calculated, performing a process of the scan speed determining step on the corrected scan speed, and repeating the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range; and an ion implanting step of implanting ions into the substrate in accordance with the practical scan number and the practical scan speed. An ion implantation apparatus according to a second aspect of the invention is characterized in that the apparatus comprises a controlling device having a function of performing: (a) a scan speed calculating process of setting an initial value of a scan number of the substrate to 1, and calculating a scan speed of the substrate by using a beam current of the ion beam, and a dose amount to the substrate, and the initial value of a scan number of the substrate; (b) a scan speed determining process of: determining whether the scan speed of the substrate is within a predetermined allowable scan speed range or not; if the scan speed is within the allowable scan speed range, setting the current scan number and the current scan speed as a practical scan number and a practical scan speed, respectively; if the scan speed is higher than an upper limit of the allowable scan speed range, aborting a process of obtaining the practical scan number and the practical scan speed; and, if the scan speed is lower than a lower limit of the allowable scan speed range, incrementing the scan number by one to calculate a corrected scan number; (c) a corrected-scan speed calculating process of, when the corrected scan number is calculated, calculating a corrected scan speed by using the corrected scan number, the beam current, and the dose amount; (d) a repeating process of, when the corrected scan speed is calculated, performing a process of the scan speed determining step on the corrected scan speed, and repeating the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range; and (e) an ion implanting process of implanting ions into the substrate in accordance with the practical scan number and the practical scan speed. In the ion implantation method or apparatus, under conditions that the scan speed of the substrate is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate in the vicinities of scan ends can be suppressed. In an ion implantation method and apparatus according to a third aspect of the invention, the scan speed determining step may include a scan number determining step of, if the scan speed is within the allowable scan speed range, determining whether the current scan number is even or odd; if the current scan number is even, setting the current scan number and the current scan speed as the practical scan number and the practical scan speed, respectively; and, if the current scan number is odd, incrementing the scan number by one to calculate a corrected scan number, and the repeating step may repeat the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range and the corrected scan number becomes even. In an ion implantation method and apparatus according to a fourth aspect of the invention, the function of performing (b) the scan speed determining process may include a scan number determining process of, if the scan speed is within the allowable scan speed range, determining whether the current scan number is even or odd; if the current scan number is even, setting the current scan number and the current scan speed as the practical scan number and the practical scan speed, respectively; and, if the current scan number is odd, incrementing the scan number by one to calculate a corrected scan number, and the function of performing (d) the repeating process may repeat the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range and the corrected scan number becomes even. In an ion implantation method and apparatus according to fifth and sixth aspects of the invention, in the scan speed calculating step or the scan speed calculating process, the initial value of the scan number of the substrate may be set to 2 in place of 1. In an ion implantation method and apparatus according to seventh and eighth aspects of the invention, in the scan speed calculating step or the scan speed calculating process, the initial value of the scan number of the substrate may be set to 2 in place of 1, and, in the scan speed determining step or the scan speed determining process, the scan number may be incremented by 2 in place of one to calculate the corrected scan number. An ion implantation method according to a ninth aspect of the invention is a method of implanting ions into a substrate using both a ribbon-like ion beam in which, with or without performing an X direction sweep, a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, a mechanical scan of the substrate in a direction intersecting with a principal face of the ion beam, and performance of ion implantation while, during a period when the ion beam does not impinge on the substrate, rotating the substrate by a step of 360/m deg. about a center portion of the substrate, and dividing one rotation of the substrate into a plurality m of implanting steps, the method comprising: a scan speed calculating step of setting an initial value of a scan number of the substrate per implanting step to 1, and calculating a scan speed of the substrate by using a beam current of the ion beam, a dose amount to the substrate, a implanting step number, and the initial value of the scan number of the substrate, and; a scan speed determining step of: determining whether the scan speed of the substrate is within a predetermined allowable scan speed range or not; if the scan speed is within the allowable scan speed range, setting the current scan number per implanting step, and the current scan speed as a practical scan number per implanting step, and a practical scan speed, respectively; if the scan speed is higher than an upper limit of the allowable scan speed range, aborting a process of obtaining the practical scan number per implanting step, and the practical scan speed; and, if the scan speed is lower than a lower limit of the allowable scan speed range, incrementing the scan number per implanting step by one to calculate a corrected scan number per implanting step; a corrected-scan speed calculating step of, when the corrected scan number per implanting step is calculated, calculating a corrected scan speed, by using the corrected scan number per implanting step, the beam current, the dose amount, and the implanting step number; a repeating step of, when the corrected scan speed is calculated, performing a process of the scan speed determining step on the corrected scan speed, and repeating the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range; and an ion implanting step of implanting ions into the substrate in accordance with the practical scan number per implanting step and the practical scan speed. An ion implantation apparatus according to a tenth aspect of the invention is an apparatus for implanting ions into a substrate using both a ribbon-like ion beam in which, with or without performing an X direction sweep, a dimension in an X direction is larger than a dimension in a Y direction that is orthogonal to the X direction, a mechanical scan of the substrate in a direction intersecting with a principal face of the ion beam, and performance of ion implantation while, during a period when the ion beam does not impinge on the substrate, rotating the substrate by a step of 360/m deg. about a center portion of the substrate, and dividing one rotation of the substrate into a plurality m of implanting steps, the apparatus comprising: a controlling device having a function of performing: (a) a scan speed calculating process of setting an initial value of a scan number of the substrate per implanting step to 1, and calculating a scan speed of the substrate by using a beam current of the ion beam, a dose amount to the substrate, a implanting step number and the initial value of the scan number of the substrate per implanting step; (b) a scan speed determining process of: determining whether the scan speed of the substrate is within a predetermined allowable scan speed range or not; if the scan speed is within the allowable scan speed range, setting the current scan number per implanting step, and the current scan speed as a practical scan number per implanting step, and a practical scan speed, respectively; if the scan speed is higher than an upper limit of the allowable scan speed range, aborting a process of obtaining the practical scan number per implanting step, and the practical scan speed; and, if the scan speed is lower than a lower limit of the allowable scan speed range, incrementing the scan number per implanting step by one to calculate a corrected scan number per implanting step; (c) a corrected-scan speed calculating process of, when the corrected scan number per implanting step is calculated, calculating a corrected scan speed, by using the corrected scan number per implanting step, the beam current, the dose amount, and the implanting step number; (d) a repeating process of, when the corrected scan speed is calculated, performing a process of the scan speed determining step on the corrected scan speed, and repeating the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range; and (e) an ion implanting process of implanting ions into the substrate in accordance with the practical scan number per implanting step and the practical scan speed. In the ion implantation method or apparatus, under conditions that the scan speed of the substrate is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number per implanting step which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate in the vicinities of scan ends can be suppressed. In an ion implantation method and apparatus according to an eleventh and twelfth aspect of the invention, the implanting step number may be even-numbered. According to the inventions set forth in the first, second, fifth and sixth aspects, under conditions that the scan speed of the substrate is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times in the vicinities of scan ends can be suppressed, and the throughput can be improved. According to the inventions set forth in the third, fourth, seventh and eighth aspects, under conditions that the scan speed of the substrate is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate in the vicinities of scan ends can be suppressed, and the throughput can be improved. Moreover, the practical scan number of the substrate can be surely even-numbered. As a result, a time loss due to the moving time for one extra scan in the case where the scan number is odd can be eliminated. Also from this viewpoint, therefore, the throughput can be improved. According to the inventions set forth in the ninth and tenth aspects, under conditions that the scan speed of the substrate is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number per implanting step which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate in the vicinities of scan ends can be suppressed, and the throughput can be improved. According to the inventions set forth in the eleventh and twelfth aspects, the implanting step number is even, and hence the total scan number can be surely even-numbered irrespective of whether the scan number per implanting step is odd or even. As a result, a time loss due to the moving time of one extra scan in the case where the total scan number is odd can be eliminated. Also from this viewpoint, therefore, the throughput can be improved. Other features and advantages may be apparent from the following detailed description, the accompanying drawings and the claims. FIG. 1 is a side view showing an embodiment of an ion implantation apparatus for implementing the ion implantation method of the invention. The portions which are identical or corresponding to those of the related-art example shown in FIGS. 9 to 11 are denoted by the same reference numerals, and, in the following description, emphasis is placed on differences from the related-art example. The ion implantation apparatus includes, in addition to the configuration of the above-described related-art ion implantation apparatus, a controlling device 30 having a function of performing a calculation control which will be described later, and a beam current measuring device 32 which measures the beam current of the ion beam 4. The beam current of the ion beam 4, and the dose amount of the substrate 2 are given to the controlling device 30. In the embodiment, more specifically, a measurement value which is measured by the beam current measuring device 32 is given as the beam current of the ion beam 4. The dose amount is given as a preset value. The beam current measuring device 32 is, for example, a Faraday cup, and receives the ion beam 4 which is conducting the ion implantation into the substrate 2, at a position where the device does not interfere with the ion implantation into the substrate 2 (for example, in the vicinity of one end in the X direction of the ribbon-like ion beam 4), and measures the beam current of the ion beam. The controlling device 30 has a function of controlling the substrate driving device 10, specifically, the scanning device 18, twist motor 14, and tilt motor 16 which constitute the substrate driving device. More specifically, the controlling device 30 performs the calculation control which will be described below, to implement an ion implantation method which will be described below. An example will be described with reference to FIG. 2. As described above, the beam current of the ion beam 4, and the dose amount of the substrate 2 are given to the controlling device 30 (step 100). Furthermore, 1 is given as an initial value of the scan number (for example, 1 is set, step 101). In accordance with Expression 2 described above or an expression which is mathematically equivalent thereto, for example, the speed of the scan of the substrate 2 which is performed for realizing the dose amount by the substrate driving device 10 (more specifically, the scanning device 18) is calculated by using these values (step 102). The steps 100 to 102 constitute a scan speed calculating step. Next, it is determined whether the scan speed of the substrate 2, more specifically, the above-calculated scan speed or a corrected scan speed which will be described later is within a predetermined allowable scan speed range or not (step 103). For example, the allowable scan speed range is a speed range which can be realized by the scanning device 18, such as a range from 100 mm/sec. to 320 mm/sec. If the scan speed is within the allowable scan speed range, the current scan number and the current scan speed are set as a practical scan number (namely, which is to be practically used, the same shall apply hereinafter) and a practical scan speed, respectively (step 104). If the scan speed is higher than the upper limit of the allowable scan speed range, the process of obtaining the practical scan number and the practical scan speed is aborted (step 105), because, even when the scan number is later increased, only a situation where the scan speed is increased is produced, and therefore the scan speed cannot be caused to be within the allowable scan speed range. In this case, for example, the implantation conditions (the beam current and the dose amount) are changed, and the process is are again performed with starting from step 100. If the scan speed is lower than the lower limit of the allowable scan speed range, the scan number is incremented by 1 to calculate the corrected scan number (step 106). The steps 103 to 106 constitute a scan speed determining step. In the case where the corrected scan number is calculated, in accordance with Expression 2 described above or an expression which is mathematically equivalent thereto, for example, the corrected scan speed of the substrate 2 for realizing the dose amount is calculated by using the corrected scan number, the beam current, and the dose amount (step 107). This step is a corrected-scan speed calculating step. In the case where the corrected scan speed is calculated, then, the process returns to step 103, and the scan speed determining step is performed on the corrected scan speed to repeat the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range. This step is a repeating step. Therefore, the scan number is incremented by a step of 1 from the initial value of 1. As a result, the practical scan number and the practical scan speed are obtained (step 104), and hence ion implantation is performed on the substrate 2 in accordance with the practical scan number and the practical scan speed (step 108). This step is an ion implanting step. Preferably, the beam current of the ion beam 4 is not a preset value but a value measured by the beam current measuring device 32 as in the embodiment. According to the configuration, even when the beam current fluctuates during ion implantation into the single substrate 2, a control of changing the scan speed in direct proportion to the beam current can be performed, so that uniform ion implantation in the Y direction can be realized without being affected by the fluctuation of the beam current. The control in which the scan speed is in direct proportion to the beam current as described above is disclosed also in, for example, JP-A-3-114128 (see the upper left column of page 2) and Japanese Patent No. 3,692,999 (see Paragraph [0037]) . In the ion implantation in step 108, there are two cases: (a) the ion implantation into the single substrate 2 is always performed at the scan speed; and (b) the ion implantation is performed with using also a control in which, while using the practical scan speed as a reference, the scan speed is in direct proportion to the beam current as described above during ion implantation into the single substrate 2. The term “in accordance with the practical scan speed” in step 108 described above is used in the meaning that both the cases (a) and (b) are included. This is applicable also to ion implantation (steps 108, 118) in other embodiments which will be described later. In the embodiment, the controlling device 30 has a function of performing: a scan speed calculating process which is identical in content to the scan speed calculating step; a scan speed determining process which is identical in content to the scan speed determining step; a corrected-scan speed calculating process which is identical in content to the corrected-scan speed calculating step; a repeating process which is identical in content to the repeating step; and an ion implanting process which is identical in content to the ion implanting step. The controlling device has a further function of performing a control in which the scan speed is in direct proportion to the beam current as described above during ion implantation into the substrate 2. In the ion implantation method (apparatus) of the embodiment, under conditions that the scan speed of the substrate 2 is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses (for example, about 0.4 to 0.5 sec. per scan) which mainly consist of deceleration and acceleration times of the substrate 2 in the vicinities of scan ends can be suppressed, and the throughput can be improved. When the scan number is reduced, the scan speed is lowered in order to realize the same dose amount (see Expression 2), but the implantation time of the substrate 2 is not changed. From this point of view, the throughput is not reduced. This will be described by way of an example. It is assumed that an implantation time which is required for implanting a desired dose amount at a certain beam current is, for example, 6 sec. Even when implantation is performed in six split implantation times of 1 sec., or in three split implantation times of 2 sec., the total implantation time is 6 sec. or unchanged. Examples of results of the measurement of the throughput will be briefly described. When the scan number was reduced from nine to three, the throughput was improved by about 7%, and, when the scan number was reduced from seven to three, the throughput was improved by about 5%. Next, another embodiment will be described with reference to FIGS. 3 to 5 and 8. The portions which are identical or corresponding to those of FIG. 2 are denoted by the same reference numerals, and, in the following description, emphasis is placed on differences from FIG. 2. In the embodiment shown in FIG. 3, step 109 is added to the flowchart of FIG. 2. If it is determined in step 103 described above that the scan speed of the substrate 2 is within the predetermined allowable scan speed range, it is determined whether the scan number is even or odd (step 109). If the scan number is even, the control proceeds to step 104 described above, and the scan number and the scan speed are set as the practical scan number and the practical scan speed, respectively. If the scan number is odd, the control proceeds to step 106 described above, and the scan number is incremented by 1 to calculate the corrected scan number. In the embodiment shown in FIG. 3, in place of the above-described scan speed determining step, there is a scan speed determining step including a scan number determining step, which is configured by steps 103 to 106 and 109 described above. In the embodiment shown in FIG. 3, the controlling device 30 has a function of performing a scan speed determining step including a scan number determining step which is identical in content to the scan speed determining step including the scan number determining step, in place of the scan speed determining process. In the ion implantation method (apparatus) of the embodiment shown in FIG. 3, similarly with the embodiment shown in FIG. 2, under conditions that the scan speed of the substrate 2 is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate 2 in the vicinities of scan ends can be suppressed, and the throughput can be improved. Moreover, the practical scan number of the substrate 2 can be surely even-numbered. As a result, a time loss due to the moving time of one extra scan in the case where the scan number is odd can be eliminated. Also from this viewpoint, the throughput can be improved. Effects due to the configuration where the practical scan number is even-numbered will be described in more detail. In the case where the scan number of the substrate 2 is odd, as indicated by the dash-dot-dot line in FIG. 1, the substrate 2, the holder 12, and the like at the end of the ion implantation into the substrate 2 are located in the other end 22 of the scan. As described above, the position of replacement of the substrate 2 with respect to the holder 12 is in the end 20 of the scan (see also FIG. 10). After the ion implantation, therefore, the substrate 2, the holder 12, and the like must be moved (in this example, lowered) by a distance corresponding to one scan. The moving time for the one scan is extra and becomes a time loss. For example, the time loss per substrate is about 1 to 1.6 sec. The time loss causes the throughput of the ion implantation to be lowered. By contrast, in the case where the practical scan number is surely even-numbered, the time loss due to the moving time for the one scan can be eliminated, and therefore the throughput can be improved. Results of measurements of the throughput in the case where both the reduction of the scan number and the even numbering are used as in the embodiment shown in FIG. 3 will be briefly exemplified. When the scan number was reduced from nine to four, the throughput was improved by about 10%, and, when the scan number was reduced from seven to four, the throughput was improved by about 10%. In an embodiment shown in FIG. 4, the initial value in step 101 in the flowchart of FIG. 2 is set to 2. Also in the ion implantation method (apparatus) of the embodiment, as described in the embodiment shown in FIG. 2, under conditions that the scan speed of the substrate 2 is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate 2 in the vicinities of scan ends can be suppressed, and the throughput can be improved. An embodiment shown in FIG. 5 is configured so that, in step 106 of the flowchart of FIG. 4, the corrected scan number is calculated while the scan number is incremented by 2 in place of the increment of one. Also in the ion implantation method (apparatus) of the embodiment, as described in the embodiment shown in FIG. 4, under conditions that the scan speed of the substrate 2 is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate 2 in the vicinities of scan ends can be suppressed, and the throughput can be improved. Moreover, the practical scan number can be surely even-numbered. As described in the embodiment shown in FIG. 3, therefore, a time loss due to the moving time of one extra scan in the case where the scan number is odd can be eliminated, and the throughput can be improved. FIG. 6 shows an example of the practical scan number and practical scan speed of the substrate 2 in the case where the ion implantation method (apparatus) of the embodiment shown in FIG. 5 is employed. FIG. 6 is a graph which, in a similar manner as FIG. 12, shows transitions of the practical scan number and the practical scan speed in the case where the dose amount is fixed and the beam current is reduced. The dose amount is equal to that in the case of FIG. 12. The maximum scan speed is equal to the reference scan speed in the case of FIG. 12. In accordance with reduction of the beam current, the scan number is increased. However, it is seen that the increase is suppressed to a very small degree as compared with the case of FIG. 12, and the scan number is always even. Also in the case where the beam current is fixed and the dose amount is increased, a similar tendency is observed, and the scan number is increased while maintained to be even. Next, an embodiment in which step implantation is performed will be described. In step implantation, ion implantation is performed while, during a period when the ion beam 4 does not impinge on the substrate 2, the substrate 2 is rotated by a step of 360/m deg. about the center portion 2a of the substrate 2 in, for example, the direction of the arrow A (or the opposite direction), and one rotation of the substrate is divided into a plurality (namely, an integer of two or more) m of implanting steps. Namely, m is the implanting step number. The implantation method is also called step rotation implantation. In the embodiment, the twist motor 14 of the substrate driving device 10 is used for the rotation of the substrate 2. The scan number n per implanting step is an integer of one or more. Therefore, the total scan number N is expressed by the following expression.N=mn[time] [Expression 3] FIG. 7 shows an example in the case where the implanting step number m is 2, the scan number n per implanting step is 2, and the total scan number N is 4. In this case, two scans or scans S1 and S2 are performed on the substrate 2 in a first implanting step ((A) of FIG. 7), the substrate 2 is then rotated by 180 (=360/2) deg. ((B) of FIG. 7), and two scans or scans S3 and S4 are then performed on the substrate 2 in a second implanting step ((C) of FIG. 7). In the embodiment, the scans S1 to S4 are performed by using the scanning device 18 of the substrate driving device 10. The detail of this is as described above. FIG. 8 shows a flowchart in the case where the step implantation is performed. The flowchart will be described while emphasis is placed on differences from FIG. 2. In the embodiment, the beam current of the ion beam 4, the dose amount to the substrate 2, and the implanting step number are given to the controlling device 30 (step 110). Furthermore, 1 is given as an initial value of the scan number per implanting step (for example, 1 is set, step 111). In accordance with Expression 4 described below or an expression which is mathematically equivalent thereto, for example, the speed of the scan of the substrate 2 which is performed for realizing the dose amount by the substrate driving device 10 (more specifically, the scanning device 18) is calculated by using these values (step 112). The steps 110 to 112 constitute the scan speed calculating step. scan speed [ cm / sec ] = implanting step number × beam current × 10 - 6 [ C / sec ] × scan number per implanting step [ time ] dose amount [ ions / cm 2 ] × elementary electric charge [ C ] × coefficient [ Expression 4 ] Steps 113, 115 are substantially identical with the steps 103, 105 of FIG. 2, respectively. If the scan speed is within the allowable scan speed range, the current scan number per implanting step and the current scan speed are set as a practical scan number per implanting step and a practical scan speed, respectively (step 114). If the scan speed is lower than the limit of the allowable scan speed range, the scan number per implanting step is incremented by 1 to calculate the corrected scan number per implanting step (step 116). The steps 113 to 116 constitute the scan speed determining step. In the case where the corrected scan number per implanting step is calculated, in accordance with Expression 4 described above or an expression which is mathematically equivalent thereto, for example, the corrected scan speed of the substrate 2 for realizing the dose amount is calculated by using the corrected scan number per implanting step, the beam current, and the dose amount (step 117). This step is the corrected-scan speed calculating step. In the case where the corrected scan speed is calculated, then, the process returns to step 113, and the scan speed determining step is performed on the corrected scan speed to repeat the scan speed determining step and the corrected-scan speed calculating step until the corrected scan speed is within the allowable scan speed range. This step is the repeating step. Therefore, the scan number per implanting step is incremented by a step of 1 from the initial value of 1. As a result, the practical scan number per implanting step and the practical scan speed are obtained (step 114), and hence ion implantation is performed on the substrate 2 in accordance with the number and the speed (step 118). This step is the ion implanting step. In the embodiment, the controlling device 30 has a function of performing: a scan speed calculating process which is identical in content to the scan speed calculating step; a scan speed determining process which is identical in content to the scan speed determining step; a corrected-scan speed calculating process which is identical in content to the corrected-scan speed calculating step; a repeating process which is identical in content to the repeating step; and an ion implanting process which is identical in content to the ion implanting step. The controlling device has a further function of performing a control in which the scan speed is in direct proportion to the beam current as described above during ion implantation into the substrate 2. In the ion implantation method (apparatus) of the embodiment, under conditions that the scan speed of the substrate is within the predetermined allowable scan speed range, ion implantation can be performed in a scan number per implanting step which is as small as possible. Therefore, accumulation of time losses which mainly consist of deceleration and acceleration times of the substrate 2 in the vicinities of scan ends can be suppressed, and the throughput can be improved. The implanting step number which is given in step 110 may be even-numbered. According to the configuration, the total scan number can be surely even-numbered irrespective of whether the scan number per implanting step is odd or even. As a result, a time loss due to the moving time of the above-described one extra scan in the case where the total scan number is odd can be eliminated. Also from this viewpoint, the throughput can be improved. The detail of the scan process is as described above in the embodiment of FIG. 3. |
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055531083 | description | BEST MODE FOR CARRYING OUT THE INVENTION In accordance with this invention and as already noted above, the upper portion of the conventional water rod W including the upper end plug 12, extension tube 20, and hold down spring 23 can be eliminated, and the upper end of the water rod left open. This arrangement, however, requires other means for securing the water rod and for preventing axial movement and rotation thereof. The invention here is easily understood from FIG. 3 which illustrates the invention as applied to a simplified lower tie plate. Specifically, a lower end plug 50 is threaded as shown at 52, and is provided with a transversely oriented, diametrical slot 54 extending from the free end of the plug, axially over part of the length of the plug. At the same time, a boss 56 in the lower tie plate 57 which receives the lower end plug 50 is threaded on its interior surface, as shown at 58, and is also formed with a pair of aligned, diametrically opposed slots 60, 62. It will be appreciated that when the lower end plug 50 is threaded into the boss 56, slots 54, 60 and 62 may be radially aligned so that a key 64 can be inserted through the slots 54, 60 and 62 to lock the water rod W in the correct angular position, and to prevent further rotation thereof in either direction. By thus preventing rotational movement of the threaded end plug, axial movement of the plug is also precluded. The key 64 may then be attached to the boss 56 by a small weld to ensure that it does not come loose. While the key 64 is shown as having a rectangular shape to cooperate with similarly shaped slots 54, 60 and 62, it will be appreciated that other slot (or bore) and key arrangements are within the scope of this invention. Turning now to FIGS. 4-6, the invention is illustrated in its preferred form, utilizing reference numerals similar to those used in FIG. 3 for corresponding elements, but with the prefix "1" added. Thus, the lower end plug 150 is threaded at 152 and is provided with a transversely oriented diametrical slot 154 which extends axially through a substantial portion of the threaded end 152. The boss 156 in the lower tie plate 157 is adapted to receive the lower end plug 150 by reason of its threaded interior surface, illustrated at 158. The boss 156 is also formed with a pair of aligned diametrically opposed slots 160, 162 within a central thickened web area 165 of the lower tie plate 157. It is noted that the tie plate illustrated in FIG. 5 illustrates a pair of water rod bosses 156, 156A, and the boss 156A may also be modified to receive a threaded end plug of the type shown at 152. The assembly of the end plug 150 to the lower tie plate 157 is achieved as described above in conjunction with FIG. 3 via the use of a key 64 (not shown in FIGS. 4-6). While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims. |
054955114 | description | DETAILED DESCRIPTION OF INVENTION In the following the present invention will be described with reference to examples of embodiments in which the inerting elements are combined with a catalyst structure to form a protective unit. It should be reemphasized at this point that this combination of a catalyst structure and inerting elements causes additional synergetic effects and that, even though only such examples of embodiments referring to this are described, inerting elements according to the present invention can be employed independently of any catalyst structure. FIGS. 1 and 2 show a preferred embodiment of the present invention in the form of a protective unit 1, which is composed of a catalyst plate 2 and inerting elements 3 attached to it. In catalyst plate 2, for example, a carrier plate made of stainless steel is shown, which preferably will be coated on either side with a catalyst material as described, for example, in DE-A-37 25 290. Instead of this, the catalyst plate 2 could also consist entirely of the catalyst material. The inerting elements 3 are each comprised of a box-like receptacle 4 made of a grid-like material of stainless steel wire which is equipped in the region of its bottom with a flange 5 jutting outwardly for attachment to the catalyst plate 2. The box, as will be recognizable in the partially broken open sectional drawing in FIG. 2, is filled with the inerting material 9 in the form, for example, of powder, granules, crystallite or the like. The receptacle 4 of the inerting element 3, as indicated in FIG. 2, can be connected to the catalyst plate with the aid of bolts 7 and nuts or in any other suitable manner. As may be seen in FIG. 2, in the embodiment depicted the bottom of the receptacle 4 is open but a filter layer 8 some 1 mm thick is located between the surface of the catalyst plate 2 and the inerting material. An identical filter layer is provided between the walls of the receptacle 4 and the inerting material located within. The filter layer is a so-called HEPA (High Efficiency Particulate Air) filter. These are filters composed of glass wool and a binding material, which have a high resistance at higher temperatures (up to some 850.degree. C.). These filters will allow hydrogen, oxygen, steam and CO.sub.2 to pass through, but shield off aerosols and fatty particles from the inerting material and, in addition, prevent any direct contact between the catalyst plate and the inerting material which might possibly cause any undesired reactions. If the inerting material is in the form of particles before and/or after disintegration, the filter will then simultaneously prevent it from falling through the grid-like walls of the receptacle 4. One or several of these inerting elements (four are shown in FIG. 1 by way of example) is attached to the catalyst plate 2 in such a manner that some catalyst surface will still be left free for the catalytic reaction. The portion of the surface left free, as well as the number and size of the inerting elements, are determined in dependence upon the place in which the protective unit is to be deployed, more precisely, in dependence upon the supply of hydrogen to be expected and the necessary degree of inerting expected in the respective place where deployed. By increasing the height H of the inerting elements with a corresponding decrease in its breadth B, the portion of free catalyst surface can be increased. In dimensioning it thus, it has to be considered that, given sufficient porosity of the inerting material, the surface of the catalyst plate covered by the inerting elements is not entirely lost for purposes of the catalytic reaction since hydrogen and oxygen can also penetrate the inerting elements through to this surface. This will apply in even greater degree to the condition following disintegration of the inerting material since the porosity of the products remaining after disintegration, with most inerting materials that come into consideration will be much larger than that prior to disintegration. As soon as hydrogen is released into the atmosphere of the compartment containing one or several protective units 1 of this type, the, catalytic transformation of the hydrogen with the oxygen into water will occur and the catalyst plate 2 will heat up. Owing to the good heat conductivity of the catalyst plate 2, the surface areas covered with the inerting elements will also be heated up evenly regardless whether they participate themselves in the catalytic reaction or not. The resulting heat will heat up the inerting material in the receptacles 4 of the inerting elements until the disintegration of the inerting materials and inerting begins to occur through the release of CO.sub.2 and/or steam when the characteristic temperatures of reactions am reached for each respective inerting material. The convection currents forming in the vicinity of the catalyst plate having a higher temperature than the surrounding will carry along the inerting gases that form and mix them with the surrounding atmosphere. It should be emphasized at this point that although in the example of the embodiment described here the inerting elements 3 are arranged only on one side of the catalyst plate 2, a duplicate arrangement of additional inerting elements can be provided on the other side of the catalyst plate. If this is done, the inerting elements on either of the two opposing sides are preferably arranged in staggered order to one another so that a uniform distribution of heat and reaction capacity will be guaranteed for the catalyst plate. To achieve a release of inerting gases staggered in time, it is possible either to place inerting materials having different temperatures of reaction inside each individual inerting element or various inerting elements could contain inerting materials with different temperatures of reaction. In the former case the different inerting materials will be placed in layers preferably parallel to the catalyst plate so that the inerting material with the highest temperature of reaction will be located closest to the surface of the catalyst and that with the lowest temperature of reaction will be situated the most distant. In so doing, the gases forming at the time the inerting material directly adjacent to the surface of the catalyst disintegrates and, by pressing outwardly, will cause the materials situated farther outward to heat up and will cause them to disintegrate, if this has not already happened. The temperature of reaction of the inerting material situated in the immediate vicinity of the catalyst structure should lie at around 200.degree. to 450.degree. C., and preferably in the range of 300.degree. to 350.degree. C., so that the synergistic effect between these two types of protective units will be optimized. Since, as pointed out above, inerting materials are available with a temperature of reaction in the vicinity of 100.degree. C., their use would cause a release of inerting gases at the beginning, even before any notable heating-up of the catalyst plate 2 had occurred. As the catalyst plate 2 heats up and passes on a part of this heat to the inerting elements, the inerting materials with higher temperatures of reaction would start to take their effect as the respective temperature of reaction is reached for each one. The transfer of heat from the catalyst plate 2 to the inerting elements 3 has the beneficial side-effect of limiting the increase in temperature of the catalyst plate. This prevents the catalyst plate from reaching a temperature that in turn could trigger an ignition, at least until the complete inerting process caused by the inerting elements has occurred. If it is anticipated that a great amount of hydrogen will be released within one of the compartments protected by the device in accordance with the present invention, it is preferable to employ for at least part of the inerting materials such a material that has a temperature of reaction of around 100.degree. C. In this manner the desired inerting effect will occur at the beginning phase of an accident, to remove the danger of flame propagation and detonation. Hydrogen is removed by the catalytic transformation before a concentration of hydrogen required for deflagration is reached. FIG. 3 shows a perspective view corresponding to that in FIG. 2 of a variation of an embodiment which differs from the one previously described in that the inerting elements are attached to the catalyst plate 2 with the intervening layer of a bottom 10 and insertion of spacing disks 11 in such a manner that an interspace is created between the inerting element and the surface of the catalyst plate. By means of this interspace, the surrounding atmosphere has free access to the catalyst surface which will be even more enhanced by the convection currents forming as the catalyst plate heats up. The bottom plate 10 and the spacing disks 11 will be made preferably of metal with good heat conductivity to guarantee the desired transfer of heat from the catalyst plate to the inerting element. Provision can be made for holes in the bottom plate 10 through which the inerting gases forming in the region of the bottom of the inerting element 3 can escape. FIG. 4 is a perspective view illustrating of a further embodiment of the present invention in which a catalyst structure and the inerting elements are arranged in the form of a concentric cylinder. A first inerting element 14 is located in the center with a hollow cylinder 15 made of a grid-like material and containing the inerting material. The inerting element 14 is surrounded concentrically by the catalyst structure 16 in the form of a cylindrical casing made of stainless sheet steel coated with the catalyst material. The catalyst structure 16 is surrounded by a second inerting element 17 with an outer cylindrical casing 18 made of grid-like material. The latter is covered on its outer side with a filter layer 19 composed of the HEPA filter material mentioned above. If the cylinder casing of the catalyst structure 16 is coated with the catalyst material only on the inside, its outer surface can simultaneously form the inner border area of the second inerting element, as shown in FIG. 4. If, however, the catalyst cylinder is also coated on the outside, depending on what catalyst surface is desired, then the second inerting element could usefully have an additional inner cylinder casing, not shown in FIG. 4, made of grid-like material which would be situated at a suitable interval from the catalyst cylinder casing to allow for the adequate flow of gases. A bottom 20, also made of a grid-like material, closes off the second inerting element below and serves as a support for the inerting material. The appropriate means for making the mechanical connection between the individual parts of the structure shown in FIG. 4 are not included in this illustration for simplicity's sake. The shaft-like structure of the arrangement seen in FIG. 4 will cause an increased convection current because of a chimney effect, with the result that the inerting gases forming will be quickly distributed. Although not shown in FIG. 4, filter layers similar to the filter layers 8 shown in FIG. 2 and 3 could be placed here between the inerting material and each respective grid enclosing it. FIG. 5 is a diagram relating to a compartment capacity of 50 m.sup.3 which shows, for four different inerting materials, the calculated steam content that can be achieved as a function of the mass of the inerting material. FIG. 6 is a similar diagram which shows, for four different inerting materials, the calculated CO.sub.2 content that can be achieved as a function of the mass. Reference is now made to FIG. 5. The four inerting materials considered here by way of example release steam at different temperatures of reaction (temperatures of disintegration); thus, the temperature of reaction (temperature of disintegration) of KAl(SO.sub.4).sub.1.12H.sub.2 O is only slightly higher than 100.degree. C. while Na.sub.2 B.sub.4 O.sub.7.10H.sub.2 O needs to be raised to a temperature of reaction of 350.degree. C. The combination of various inerting materials with different temperatures of reaction within various inerting elements causes, as pointed out above, the release of steam staggered in time in accordance with the increase in temperature. If the lowest of the different temperatures of reaction of the various inerting materials lies at about 100.degree. C., which is to say at a temperature which could even be reached without any heating-up caused by catalytic reaction, then the pertinent inerting material will not take away any heat from the catalyst plate; thus, the catalyst plate will then quickly reach a higher temperature that will contribute significantly to the catalytic reaction. The more hydrogen that is catalytically converted per unit time, the more intensely the concentration of hydrogen will be lowered but more inerting will also be achieved because of the steam generated by the catalytic reaction. It is immaterial for inerting by steam whether the steam comes as the result of catalytic transformation of hydrogen or from the inerting material. The steam will need a certain amount of time to reach a cold wall at which it condenses. A high concentration of steam will establish itself, despite its being constantly condensed at the cold walls, because it is steadily produced through catalytic reaction and through the inerting elements. This condensation contributes to avoiding toe, great an increase in pressure caused by the steam being released. After the catalytic reaction is completed, because of the removal of the hydrogen, it is possible that a global relief in pressure will occur; therefore, an active reduction in the pressure of the atmosphere in the reactor containment can possibly be avoided. It may be seen in FIG. 5 that as little as 10 kg of the inerting materials described can contribute 12 to 15% by volume of the steam (related to a compartment capacity of 50 m.sup.3). In FIG. 6 similar conditions prevail for the production of CO.sub.2. At this point it should be pointed out that the chemical substances used as examples of inerting materials have a 40 to 60% portion of water of crystallization or CO.sub.2. If inerting materials with higher portions are employed, it will be possible to work with correspondingly lower masses. FIG. 7 illustrates an embodiment of the invention in which various materials used for releasing an inerting gas or inerting gas mixture are employed to react with one another. The embodiment depicted is suitable for cases in which a solid material and a liquid are intended to react with one another. Located in the lower region of a frame 20 is a tub 21 made of a suitable material in which hydrochloric acid, for example, may be placed. A wire basket 22 made of stainless steel is suspended above the tub 21. Calcium carbonate (in this example) is placed in the wire basket. The suspension 23 of the wire basket is constituted so that it will release when reaching a certain pre-determined temperature, whereupon the wire basket will drop into the tub 21 containing the hydrochloric acid. The size of the mesh of the wire basket will be established such that the calcium carbonate with which it is filled will not drop out, but also such that the hydrochloric acid, once the wire basket drops into the tub, can freely penetrate into it and the desired reaction will take place. The suspension can contain, for example, a soldered point with a solder having a melting temperature equal to the temperature desired for the inerting gas mixture to be released. The above embodiment may also be combined in an advantageous manner with a catalyst structure; however, this does not appear in the illustrations. For this purpose it would be necessary only to establish a connection with good heat conductivity between the suspension 23 and a catalyst structure located in close proximity. The above described embodiments of the present invention represent a large variety of forms in which the present invention can be realized. What is important is that the chemical substance(s) employed for passive inerting according to the present invention will be situated in the compartment to be protected in a manner that permits the free exchange of gas with the atmosphere of the compartment. A housing with gas-permeable walls can hold together both the substance prior to its disintegration or the reaction and also the product of the reaction that remains. Wherever an inerting element in accordance with the present invention is used in conjunction with a catalyst structure, care must be taken that the catalytic recombination is not adversely affected and that the desired transport of heat to the inerting element can take place. In addition, it must be guaranteed that once the inerting gas has been released by the inerting element, its remaining product of reaction (which accordingly should preferably not be liquid) is kept away from the catalyst structure to not impede its effect. |
summary | ||
051456397 | summary | BACKGROUND OF THE INVENTION The present invention relates to nuclear reactor plants and, more particularly, to isolation condensers for such plants. A major objective of the present invention is to provide an a simpler and more compact isolation condenser characterized by improved flow stability. Fission reactors rely on fissioning of fissile atoms such as uranium isotopes (U233, U235) and plutonium isotopes (Pu239, Pu241). Upon absorption of a neutron, a fissile atom can disintegrate, yielding atoms of lower atomic weight and high kinetic energy along with several high-energy neutrons. The kinetic energy of the fission products is quickly dissipated as heat, which is the primary energy product of nuclear reactors. Some of the neutrons released during disintegration can be absorbed by other fissile atoms, causing a chain reaction of disintegration and heat generation. The fissile atoms in nuclear reactors are arranged so that the chain reaction can be self-sustaining. Dual-phase reactors store heat generated by the core primarily in the form a phase conversion of a heat transfer medium from a liquid phase to a vapor phase. The vapor phase can used to physically transfer stored heat to a turbine and generator, which are driven to produce electricity. Condensate from the turbine can be returned to the reactor, merging with recirculating liquid for further heat transfer and cooling. Dual-phase reactors are contrasted with single-phase reactors, which store energy primarily in the form of elevated temperatures of a liquid heat-transfer medium. Pressurized water reactors (PWRs) are considered single-phase in that the reactor coolant is maintained in a liquid state, although heat from the pressurized water is used to boil a secondary coolant to drive a turbine. The primary example of a dual-phase reactor is a boiling-water reactor (BWR). The following discussion relating to BWRs is readily generalizable to other dual-phase reactors. Modern BWRs provide for the removal of reactor decay heat from a reactor pressure vessel in the event the turbine becomes isolated from the reactor. During a turbine shutdown, a valve on the main steam line is closed preventing steam from reaching the turbine. Even after the reactor is shut down by fully inserting the control rods, decay heat continues to be generated for a period of days. This heat generates steam, which if left to accumulate in the reactor pressure vessel, could exceed the vessel's pressure-bearing specifications, potentially inducing a breach. An isolation condenser is one type of system designed to handle steam during turbine isolation to avoid excessive pressure accumulation. A typical isolation condenser includes an upper distributor chamber and a lower collector chamber. The chambers are immersed in the water of a condenser pool. The chambers are coupled via an array of vertical tubes which extend therebetween and through intermediate pool water. During isolation, steam is conveyed to the distributor chamber. The steam is forced through the tubes, which through heat exchange with the condenser pool, condense the steam so that water flows into the collector chamber. A drain conduit coupled to the outlet chamber conveys the condensate to the reactor to replenish its coolant supply. The performance of such a condenser can be impaired when the condenser pool has been heated to saturation. At that point, steam generated in the pool can insulate the heat-exchanger tubes, limiting further heat transfer and causing thermal cycling in the manifold. The thermal cycling can stress the condenser, impairing its structural integrity and inducing pool-side flow instability. Other problems with such a conventional isolation condenser concern the amount of material required to ensure the distributor and collector can withstand the large pressure differentials that can develop between their interiors and the condenser pool. Pressure differentials of up to about 1250 pounds per square inch must be accommodated. The relatively flat boundaries, including the tube sheets, of the disk-shaped distributor and collector require considerable thickness to withstand this pressure. The thickness not only adds bulk and mass to the condenser, but subjects it to thermal stresses due to the larger thermal gradients that thicker material can sustain. What is needed is a more compact, lightweight isolation condenser that is less subject to flow instability. In addition, the condenser should be economical to manufacture and maintain. SUMMARY OF THE INVENTION In accordance with the present invention, a dual-phase reactor plant incorporates an isolation condenser that isolates a contiguous volume that is divided by a partition into a distributor plenum and a collector plenum. These plenums are coupled by tubes of a manifold that pass through a condenser pool in which the condenser is disposed. The condenser has a base, a vertically extending sidewall, and a domeshaped top. The cross section of the sidewall is such that the area it encloses measures at least the square of one-fourth its perimeter. The plant includes a containment structure, typically of concrete, that defines a dry well, a wet well, and a condenser well. The reactor is in the dry well, which is otherwise filled with noncondensible gases. The condenser is in the condenser well, immersed in coolant. The wet well holds a suppression pool of coolant. The reactor produces vapor which drives a turbine, which in turn can be used to drive a generator to produce electricity. During normal operation, conduits convey vapor from the reactor to the turbine and condensate from the turbine to the reactor. When the turbine is decoupled from the reactor, a resulting pressure buildup triggers a relief valve that permits vapor to escape to the distributor plenum of the condenser. The vapor is distributed to the manifold tubes where they give up heat energy to the suppression pool. The loss of energy results in condensation of the vapor to liquid. The liquid flows through the sidewall into the outlet chamber. From the collector plenum, the liquid flows into the reactor through a conduit that mates with the reactor below its nominal liquid level. Noncondensible gases accumulating in the collector plenum can be conveyed by a conduit to the suppression pool. The present design provides for enhanced flow stability relative to the conventional condenser. In the latter case, steam forming on the outside of a heat-exchanger tube tends, under the influence of gravity, to flow upward. Since the tubes are vertical, the rising steam forms an insulating sheath about the tube, impairing its heat-transfer capability. Movement of steam reaching the top of a tube is then impeded by the tube plate of the distributor chamber. Thus, steam remains in the manifold vicinity, impeding the performance of the condenser and threatening its integrity. In the present design, the heat-exchange tubes extend radially from the condenser chamber. Rising steam escapes the tube at which it was generated, perhaps passes between the tubes of another array, and then rises unimpeded to the surface of the condenser pool. This relatively free movement of steam induces convection, minimizes hot spots within the pool, and further ushers steam away from the tubes. Thus, optimal heat exchange is maintained and the structural integrity of the condenser is prolonged. The radial arrangement of heat-exchanger tubes has the further advantage that longer tubes do not require a taller condenser. Because vertical constraints do not have to be divided three ways between two chambers and the tubes, the single condenser chamber can be made relatively tall, as can each of its plenums. The relatively tall collector plenum permits a relative high entrance level for a noncondensible gas outlet, providing improved separation of noncondensible gases from condensate. The novel condenser geometry provides favorable distribution of stresses induced by pressure differentials between the suppression pool and the internal volume of the condenser. The improved stress handling allows thinner boundary walls to be used with less reinforcement. This in turn reduces thermal gradients through the walls and hot spots within the walls. Reducing thermal gradients and hot spots reduces thermal stresses and prolongs the useful lifetime of the condenser. The compactness provides proportionate advantages in the size of the containment structure, which strongly impacts plant cost. In addition, the vertical design combined with compactness, along with the employment of a relatively small and light cover, minimize the requirements for overhead access. This makes maintenance more convenient and more economical. These and other features and advantages of the present invention are apparent from the description below with reference to the following drawings. |
description | The present invention relates to an in-pile structure in a reactor vessel of a nuclear power plant and, in particular, to a neutron reflector bolt fastening structure to be used when securing in position a neutron reflector within a core vessel. The present invention also relates to a fastening method for the fastening structure. FIG. 3 is a schematic longitudinal sectional view showing the general construction of an ordinary nuclear reactor. In the drawing, a reactor vessel 10 is equipped with a core vessel 12 in which a fuel assembly 14 is supported. The fuel assembly 14 in the core vessel 12 is surrounded by an upper core plate 16 on the upper side, by a lower core plate 18 on the lower side, and by a neutron reflector 20 at the periphery. Next, the assembly structure of this neutron reflector will be described with reference to FIG. 4. The neutron reflector 20 is formed by vertically stacking together eight substantially annular stage portions, which are fastened by eight tie rods 22 that are circumferentially arranged. The positioning when assembling each stage portion is effected by a positioning pin 23. The lowermost stage portion 20A of the neutron reflector 20 has at the periphery thereof four flange portions 201 (only two of which are shown in the drawing); similarly, the top stage portion 20C has four flange portions 202 at the periphery thereof. As shown in FIG. 5, the tie rods 22 extend through the neutron reflector 20 from the top stage portion 20C to the lowermost stage portion 20A. The lowermost end portions of the tie rods 22 are passed through the lower core plate 18 and screwed into a flange portion 13 formed in the core vessel 12. To the upper end portion of each tie rod 22, there is mounted a nut 24 for pressing down the upper surface of the top stage portion 20C of the neutron reflector. By turning these nuts 24, the neutron reflector 20 is tightened between them and the lower core plate 18, and is secured to the flange portion 13 of the core vessel 12 through the intermediation of the lower core plate 18. The neutron reflector 20 has a large number of flow holes for cooling, through which cooling water flows. FIG. 6 shows a structure of inlet portions of the flow holes 204 in the lowermost stage portion 20A of the neutron reflector 20. In the drawing, plugs 181 are provided in the lower core plate 18 mounted to the flange portion 13 of the core vessel 12. Cooling water flowing in through these plugs 181 passes orifices 203 provided in the lowermost stage portion 20A of the neutron reflector 20 and flows upwards through the flow holes 204. These flow holes extend through the 8-stage neutron reflector 20 from the lowermost stage portion 20A to the top stage portion 20C, so that the cooling water flowing in through the orifices 203 of the lowermost stage portion 20A rises through the flow holes 204 to flow out through the flow holes of the top stage portion 20C of the neutron reflector 20. Where the coolant water passes through the orifices 203 of the lowermost stage portion 20A, a pressure loss is generated, and, due to this pressure loss, a great lifting force is applied to the entire assembly structure of the neutron reflector 20. Most of this lifting force is generated when the cooling water passes through the orifices 203 of the lowermost stage portion 20A, and the force applied to the remaining seven stage portions, that is, from the second stage portion 20B to the top stage portion 20C is relatively small. In view of this lifting force, the eight tie rods 22 are fastened to thereby press the neutron reflector 20 against the lower core plate 18. However, in a conventional structure, in which the neutron reflector 20 is pressed against the lower core plate 18 by the eight tie rods 22, when relaxation or loosening is generated in the tie rods 22 as a result of neutron irradiation, there is a possibility of the fastening force for pressing down the neutron reflector 20 against the lifting force falling short of the required level. Thus, it is a principal object of the present invention to provide a neutron reflector bolt fastening structure capable of firmly pressing the neutron reflector against the flange portion of the core vessel through interconnection with the lower core plate even if relaxation is generated in the tie rods as a result of neutron irradiation, as well as a fastening method for the structure. According to a main aspect of the present invention, a neutron reflector bolt fastening structure is characterized by including: a neutron reflector composed of a plurality of divided stage portions and situated in a core vessel in a reactor vessel; a plurality of tie rods for fixing the neutron reflector to the core vessel; and a plurality of bolts for solely fixing the lowermost stage portion of the plurality of stage portions of the neutron reflector to the core vessel. According to another aspect of the present invention, a neutron reflector bolt fastening method is characterized by including: fixing a neutron reflector composed of a plurality of divided stage portions and situated in a core vessel in a reactor vessel to the core vessel by means of a plurality of tie rods; and fixing the lowermost stage portion of the plurality of stage portions of the neutron reflector solely to the core vessel by means of a plurality of bolts. In accordance with the present invention, the lowermost stage portion, to which most of the lifting force on the entire neutron reflector is applied, is exclusively secured to the core vessel by means of bolts other than the tie rods, whereby the initial fastening force for the lowermost stage portion of the neutron reflector becomes very large. Thus, even if relaxation is generated in the tie rods as a result of neutron irradiation, it is possible to press the neutron reflector firmly against the core vessel. Next, a preferred embodiment of the present invention will be described with reference to the accompanying drawings, in which the components which are the same as or equivalent to those of the conventional structure are indicated by the same reference numerals. FIG. 1 is a schematic diagram showing a neutron reflector bolt fastening structure according to the present invention, and FIG. 2 is an end view taken along the line A-A of FIG. 1. In these drawings, the lowermost stage portion 20A of the neutron reflector 20 has flange portions 201, which has through-holes 205 for bolts 1. Further, the lower core plate 18 situated under the neutron reflector 20 also has through-holes 182 for the bolts 1. The flange portion 13 of the core vessel 12 supporting the neutron reflector 20 and the lower core plate 18 has screw holes 131 into which bolts 1 are to be fitted. By using the bolts 1, the lowermost stage portion 20A of the neutron reflector 20 is fastened to the flange portion 13 of the core vessel 12 through the intermediation of the lower core plate 18. In this embodiment, there are provided eight holes 205, 182, and 131, and eight bolts 1 for each flange portion 201 formed on the lowermost stage portion 20A. Namely, four flange portions 201 are provided (See FIG. 4), so that the lowermost stage portion 20A of the neutron reflector 20 is fastened to the flange portion 13 by a total of 32 bolts. When fastening the neutron reflector by utilizing this neutron reflector bolt fastening structure, the entire 8-stage neutron reflector 20 is fastened to the core vessel 12 by the conventional eight tie rods 22 and, at the same time, the lowermost stage portion 20A of the neutron reflector 20 is solely fastened to the flange portion 13 of the core vessel 12 by means of a large number of bolts 1 according to the present invention. In this manner, the lowermost stage portion 20A of the neutron reflector 20, to which most of the lifting force is applied, is solely fixed to the core vessel 12 exclusively by means of a plurality of bolts which are separate from and independent of the tie rods 22, whereby it is possible to secure in position the lowermost stage portion 20A of the neutron reflector 20 with a much larger initial fastening force as compared with the force obtained from the eight tie rods 22. Thus, even if relaxation is generated in the tie rods 22 as a result of neutron irradiation, and the fastening force is reduced, it is possible to maintain a sufficient fastening force to cope with the lifting force since the initial fastening force of the large number of bolts 1 is sufficiently large. As a result, it is possible to maintain the neutron reflector 20 in a state in which it is pressed against and fixed to the flange portion 13 of the core vessel 12. Regarding the remaining seven stage portions of the neutron reflector 20, the lifting force applied thereto is relatively small as compared with that applied to the lowermost stage portion 20A, so that the fastening with the conventional eight tie rods 22 suffices. Even if relaxation is generated in the tie rods and their axial force is weakened, it is possible to maintain the requisite fastening force. In the present invention, the bolts 1 are fastened where the amount of neutron irradiation amount, so that it is possible to prevent a deterioration in fastening force due to relaxation, thus making it possible to reliably maintain the requisite fastening force. While in the above-described embodiment all the four flange portions 201 of the neutron reflector 20 are fastened to the flange portion 13 of the core vessel 12 through the intermediation of the lower core plate 18 by means of the bolts 1, it is also possible to fasten only two opposing flange portions 201. Further, while in the above-mentioned embodiment eight bolts 1 are used for each flange portion 201, the number of bolts for each flange portion may be more or less than eight as long as they provide a predetermined initial fastening force. |
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050283804 | abstract | Disclosed are a method and device for the identification of the leakiness of a neutron-capturing pencil, or control rod, of a nuclear reactor. The pencil is placed in an impervious chamber filled with an aggressive chemical solution. The solution is put under pressure in order to make it penetrate the defective pencil, then this pressure is relaxed in order to enable the solution to go out of the presumably defective pencil in the impervious chamber. The solution is analyzed in order to show up metallic salts of the constituent elements of the core of the pencil. Application is to routine checks on neutron-capturing pencils of a pressurized-water nuclear reactor. |
description | This application claims priority of German application No. 10 2008 005 069.5 filed Jan. 18, 2008, which is incorporated by reference herein in its entirety. The invention relates to a positioning device for positioning a patient in a medical diagnostic and/or therapy system, such as is used in particular in a particle therapy system for positioning a patient relative to a treatment beam. The invention also relates to a particle therapy system having a positioning device of said kind as well as to a method for operating a positioning device of said kind. Particle therapy is an established method for treating tissue, in particular tumor diseases. However, irradiation methods as used in particle therapy are also used in non-therapeutic application areas. These include, for example, research activities in the particle therapy field that are carried out on non-living phantoms or bodies, irradiation of materials, etc. Typically, in such applications, charged particles are accelerated to high energies, formed into a particle beam and guided by way of a high-energy beam transport system to one or more irradiation rooms. In one of said irradiation rooms the object that is to be irradiated is exposed to the particle beam. In this case it is essential to the success of an exposure to radiation that the object that is to be irradiated is positioned as accurately as possible relative to the particle beam. Devices are known in which the positioning of, for example, a patient is accomplished with the aid of a robot-based positioning device. For example, a patient treatment couch can be flexibly positioned relative to a particle beam by means of a multi-axis robot arm. It is therefore the object of the invention to disclose a positioning device for positioning a patient in a medical diagnostic and/or therapy device, by means of which positioning device a high level of patient safety is ensured during the positioning of the patient. It is also the object of the invention to disclose an irradiation device which ensures a high level of patient safety during the positioning of the patient as well as a method for operating a positioning device of said kind. The object of the invention is achieved by a positioning device, an irradiation device, and a method for operating a positioning device as claimed in the independent claims. Advantageous developments can be found in the features of the dependent claims. The inventive positioning device for positioning a patient in a medical diagnostic and/or therapy device comprises: a patient receiving device on which a patient can be placed, and a robot arm having a plurality of movement axes and enabling the patient receiving device to be positioned in the room,wherein the positioning device can be placed into a manual operating mode in which a position of the patient receiving device in the room can be changed manually. The invention is based on the concept that although accurate and precise positioning of a patient is possible with known positioning devices having a multi-axis robot arm, whereby the control of the positioning device is usually handled automatically by a control unit, if necessary by interaction with a user who controls the controller by means of a manual control element, in emergency situations access to a patient can be made more difficult on account of a positioning device of that kind. In an emergency situation, for example in the event of a malfunction of the positioning device and/or of the system in which the positioning device is operated, the robot arm is normally disabled so that it will no longer be possible to move the patient by way of the control unit. This is designed to ensure that in an emergency situation of said kind the robot arm will not be able to execute any erroneous movements which would put the safety of a patient at serious risk. An emergency situation of the aforesaid kind can occur, for example, if an emergency stop has been actuated in a particle therapy system, as a result of which on the one hand the particle beam and on the other hand the power supply of many components in the treatment room are switched off. Such an emergency situation can also occur, however, if a component of a particle therapy system, for example the positioning device itself, detects a deviation from a normal mode of operation and accordingly is switched to a mode in which the scope of operation of the component and/or of further components is restricted, for example as a result of the power supply being switched off. It was recognized in such instances that if an emergency situation occurs it is advantageous to at least partially switch off individual and/or multiple components, that as a consequence of doing so, however, the safety of a patient can be compromised, since a disabling of the robot arm of the positioning device may result in the patient's remaining in a position in which an intervention directed at the patient is made more difficult or even impossible. If the positioning device is used in, for example, a gantry-based irradiation room, it is possible—depending on the particular embodiment of the irradiation room—that the irradiation room will have a moving floor which is operated in order to move the gantry out of the irradiation room. In this state the irradiation room has no floor on which a user could walk without risk. If an emergency situation arises at such a time, the positioning device can, for example, block the patient in a position in which access to the patient is impossible. Even in irradiation rooms with a floor present, however, in an emergency situation the patient can be blocked in an unfavorable position of said kind in such a way that access to the patient, for example in order to enable resuscitation measures to be carried out, is impossible. A simple and effective solution to this problem is provided by means of the positioning device according to the invention. The positioning device is embodied in such a way that it can be placed into a manual operating mode in which a position of the patient receiving device in the room can be at least partially changed manually. In this way it is now possible even in emergency situations to change the position of the patient in the room manually so that a patient can be brought into a more favorable position for recovery or treatment. The patient receiving device can be embodied e.g. as a patient treatment couch or patient chair on which a patient can be positioned in a posture provided for irradiation. In an advantageous embodiment of the invention, the positioning device is embodied in such a way that a rotation of the patient receiving device about a vertical axis can be performed manually in the manual operating mode. This can be realized, for example, such that in the case of the multi-axis robot arm one axis or more than one axis, each with a vertical axis of rotation, can be unlocked or, as the case may be, released. A rotation about one vertical axis of rotation is usually sufficient to recover a patient placed on the patient receiving device from a risk situation. The unlocking of an axis (or more than one axis) can easily be realized e.g. by selective unlocking of the brakes of an axis in the manual operating mode. In one embodiment, the positioning device is embodied in such a way that the manual operating mode is activated automatically if a fault condition occurs during normal operation of the positioning device and/or the diagnostic and/or therapy system. A fault condition indicates a malfunction, so that in that case various steps are usually taken in order to ensure the safety of the patient. In a particle therapy system, for example, the particle beam can be switched off in order to stop an irradiation process. In this embodiment, in addition, the positioning device is now placed automatically into the manual operating mode in order to permit a manual movement of the patient receiving device and hence of the patient at all times in the event of a fault condition being present. In an alternative and/or additional embodiment, the positioning device has a manually actuatable switchover device. An actuation of the switchover device places the positioning device into the manual operating mode. By this means it is possible to activate the manual operating mode of the positioning device by manual intervention. In this way a patient can be brought more quickly into a desired position than would be possible, for example, in the case of a normal, automatically controlled movement of the positioning device during operation. A normal movement of the positioning device is namely often linked to the movement of other components in the treatment room, with the result that in an automatically controlled movement of the positioning device individual movements have to be coordinated with one another, an operation which often takes up a considerable amount of time. By means of a manual actuation of the switchover device it is possible when necessary to switch to the manual operating mode so that a patient can quickly be moved manually. This can be realized, for example, in that a button disposed on the positioning device is pressed, thereby, for example, unlocking a specific axis of the robot arm. In another embodiment, the switchover device can be embodied in such a way that the switchover device is disposed on one of the movement axes of the robot arm, in particular on a movement axis having a vertical axis of rotation, and an actuation of the switchover device effects a direct mechanical unlocking of said movement axis. This is particularly advantageous since in this way an unlocking of the movement axis is possible in any case, even if the control of the positioning device has failed completely. In one embodiment, in order to simplify moving of the patient receiving device, means for manually taking hold of the patient receiving device are disposed on the patient receiving device. Means of said kind can be, for example, a handle or a receiving device into which a hook disposed on a pole can be introduced. A cord that extends from the patient receiving device as far as a user enables the patient receiving device to be moved even when it is not located within immediate reach of a user. The cord can be, for example, permanently attached to the robot arm and released when necessary by pulling. The irradiation device according to the invention comprises: a particle source for generating particles, an accelerator for accelerating the particles and for providing a high-energy particle beam, an irradiation room for exposing an object that is to be irradiated to the high-energy particle beam, and a positioning device as claimed in one of claims 1 to 7 which is disposed in the irradiation room. In a special embodiment, the irradiation room includes a movable gantry, thereby enabling the particle beam to be directed from different selectable directions onto the object that is to be irradiated, as well as a removable floor in the irradiation room, in particular in the gantry area of the irradiation room. With the aid of the removable floor, additional freedom of movement can be created for moving the gantry. In a room of this kind the use of the inventive positioning device has a particularly advantageous effect since in this instance—in the case of the positioning device becoming blocked in the area of the removed floor—it would otherwise be possible to recover a patient only with very great effort. With the inventive method for operating a positioning device by means of which a patient can be positioned in a medical diagnostic and/or therapy system: a normal operating mode is provided during which the patient receiving device is positioned fully automatically at a position predefined by means of a control device, a manual operating mode is provided during which the position of the patient receiving device in the room can be changed manually, and a switchover from the normal operating mode into the manual operating mode is performed as soon as a switchover condition is present. In one embodiment, a switchover condition can automatically be present in the case of the method whenever a fault condition is detected during the normal operating mode. In this case the fault condition can, for example, characterize a malfunction of the positioning device or else also indicate a malfunction of the medical diagnostic and/or therapy system. In another embodiment, the switchover condition is present whenever a switchover device has been actuated manually. The switchover device can be, for example, an emergency stop switch, as a result of the actuation of which a series of operations is triggered, an interruption of the irradiation process, for example. The switchover from the normal operating mode into the manual operating mode is accomplished by actuation of the switchover device. In a special embodiment, the switchover device can, however, also be disposed on a movement axis of the robot arm. In particular, a direct mechanical unlocking of the movement axis of the robot arm is possible by actuating the switchover device. FIG. 1 shows a schematic overview of the layout of a particle therapy system 10. In a particle therapy system 10, a body, in particular tumor-diseased tissue, is irradiated in particular with a particle beam. Ions such as, for example, protons, pions, helium ions, carbon ions or other types of ions are principally used as particles. Particles of said kind are typically generated in a particle source 11. If, as shown in FIG. 1, two particle sources 11 are present which generate two different types of ions, it is possible to switch between said two types of ions within a short time interval. For that purpose a switching magnet 12, for example, is used which is disposed between the ion sources 11 on the one side and a pre-accelerator 13 on the other. By this means the particle therapy system 10 can be operated, for example, with protons and with carbon ions simultaneously. The ions generated by the ion source or one of the ion sources 11 and where applicable selected by means of the switching magnet 12 are accelerated to a first energy level in the pre-accelerator 13. The pre-accelerator 13 is, for example, a linear accelerator (LINAC). The particles are then fed into an accelerator 15, for example a synchrotron or cyclotron. In the accelerator 15, they are accelerated to high energies, such as are required for irradiation. After the particles have exited the accelerator 15, a high-energy beam transport system 17 guides the particle beam to one or more irradiation rooms 19. In an irradiation room 19, the accelerated particles are directed onto a body that is to be irradiated. Depending on embodiment, this is done from a fixed direction (in what are termed “fixed beam” rooms) or else from different directions by way of a rotatable gantry 21 that is movable about an axis 22. The basic layout of a particle therapy system 10 as shown with reference to FIG. 1 is typical of many particle therapy systems, but can also differ herefrom; for example, depending on the acceleration of the particles, an irradiation device does not have to be disposed as a particle therapy system. The exemplary embodiments described hereinafter can be used both in connection with the particle therapy system illustrated with reference to FIG. 1 and with other particle therapy systems or radiotherapy systems. FIG. 2 shows a perspective view of a gantry-based irradiation room 31 in which a positioning device 45 is used for positioning a patient 49. The patient 49 is irradiated with a particle beam 33 that is emitted from a particle beam emitting device 35. The particle beam emitting device 35 is disposed on a gantry (not visible in FIG. 2) which enables the particle beam emitting device 35 to be rotated (double-headed arrow 37) in the irradiation room in such a way that the particle beam 33 can be emitted from a plurality of angles. Also disposed on the particle beam emitting device 35 shown are flat-panel detectors 39 which can be used for checking the position of the patient 49 in the irradiation room 31. In order to enable the particle beam emitting device 35 to be positioned in an angular range of preferably 0° to 360°, a part of the floor 41 in the irradiation room 31 can be moved out of the irradiation room 31 (straight double-headed arrow 43). This creates space for the rotation of the gantry, with the result that the particle beam emitting device 35 can also be positioned, for example, such that the particle beam 33 is directed onto a patient 49 from below. The positioning device 45 is embodied as a multi-axis robot arm 51 by means of which a patient receiving device 47, in this case shown as a patient treatment couch, can be positioned in the irradiation room 31. Instead of a patient treatment couch 47 it is also possible to use, for example, a patient chair on which the patient 49 can be placed in a sitting posture. Not shown in the figure is a control device by means of which the positioning device 45 is operated in order, for example, to move a patient 49 automatically to a predefined position. A first movement axis 53 of the robot arm 51 enables the patient receiving device 47 to be rotated about a first vertical axis of rotation 54. A second movement axis 55 enables the patient receiving device 47 to be rotated about a second vertical axis of rotation 56. In this case the first movement axis 53 is disposed immediately beneath the patient receiving device 47, while the second movement axis 55 is disposed on the floor of the irradiation room 31. A rotation about the first vertical axis of rotation 54 only rotates the patient receiving device 47. A rotation about the second vertical axis of rotation 56 rotates the entire robot arm 51 together with the patient receiving device 47 about the second vertical axis of rotation 56. The positioning device 45 is embodied in such a way that it can be placed into a manual operating mode which permits a rotation of the patient receiving device 47 about the first vertical axis of rotation 54 and/or about the second vertical axis of rotation 56 to be performed manually. This can be accomplished for example in that a user (not shown) takes hold of a handle 65 of the patient receiving device 47 and exerts force onto the patient receiving device 47. Alternatively and/or in addition, a user can catch an eye 61 of the patient receiving device 47 by means of a pole 63 that has a hook and exert force onto the patient receiving device 47. A further possibility is to release a cord 67 disposed on the robot arm 51 and exert a pull on the patient receiving device 47 by way of the cord 67. In an emergency situation this allows the patient 49 to be moved by means of the patient receiving device 47 in such a way that he or she can be brought manually into a position in which care of the patient 49 can be carried out better and more easily. This is the case in particular when the floor 41 of the gantry-based irradiation room 31 has been moved out of the irradiation room 31, thereby making it impossible for a direct approach to be made to the patient 49 positioned in the region of the particle beam emitting device 35. In particular when an emergency situation arises, for example when an emergency stop button 57 has been pressed by means of which both an irradiation process is aborted and automatic control of the position of the particle beam emitting device 35 and of the positioning device 47 is switched off because an unforeseen event has occurred, the patient 49 can easily be recovered from a danger zone manually without its being necessary to restart the system first in order to execute an automatic control of the positioning device 47. For this purpose the first movement axis 53 and/or the second movement axis 55 of the robot arm 51 can be released, for example, automatically when the emergency stop button 57 has been pressed. Alternatively and/or in addition, a movement axis can be released by direct actuation of a lever that is disposed on the movement axis. In FIG. 2, for example, there is disposed on the second movement axis 55 a lever 59, by the actuation of which the movement axis 55 can be directly unlocked mechanically, such that a movement of the robot arm about said movement axis is made possible manually. FIG. 3 shows a schematic overview of method-related steps that are performed when operating the positioning device. The positioning device can be operated in a normal operating mode 81 in which a completely automatic positioning of a patient by means of the positioning device is performed. Manual positioning of the positioning device is not possible in the normal operating mode 81. The positioning device can also be operated in a manual operating mode 83 in which a manual positioning of the positioning device, in particular of the patient receiving device, is made possible in the room. During the operation of the positioning device it is possible to switch from the normal operating mode 81 into the manual operating mode 83 as soon as a switchover condition 85 is present. Depending on the embodiment of the positioning device, a switchover condition 85 of said kind can automatically be present when a fault condition 87 has been detected during the operation of the positioning device, the condition being caused, for example, by a malfunction of the positioning device itself or by a malfunction of the diagnostic and/or therapy system in which the positioning device is operated. A switchover condition 85 can alternatively and/or additionally be present also when a manual actuation 89 of a switchover device has been performed. A switchover device of said kind can be, for example, the emergency stop button 57 shown in FIG. 2, by the actuation of which the switchover condition has been triggered manually, as a result of which the positioning device is placed into the manual operating mode, or, for example, a lever 59 which can be pulled. |
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summary | ||
description | This application is a divisional application of U.S. patent application Ser. No. 09/946,486, filed Sep. 6, 2001, now U.S. Pat. No. 6,838,686, issued on Jan. 4, 2005. The present invention relates to an alignment method and an exposure apparatus using the method, and particularly is suitable for an alignment method in semiconductor manufacturing and method and apparatuses for manufacturing devices using it. Currently, in semiconductor manufacturing, a semiconductor device is fabricated by depositing multiple layers successively. In actual semiconductor manufacturing, a method is known wherein instead of measuring positions of alignment marks formed in a layer prior to exposure, marks are formed in the multiple layers and alignment is performed by measuring positions of the marks in multiple layers. As described in Japanese Patent Laid-Open No. 7-321012, it is suggested that when forming a layer on a substrate, the layer is formed after measuring positions of marks formed in each of at least two layers formed prior to the layer, based on measurement of the mark positions in each of said layers. In the past, there has been a problem that, in measuring alignment marks formed in each layer, the measurement accuracy is degraded due to manufacturing processes such as the physical feature and resist application condition of each alignment mark. It is an object of the present invention to provide an alignment method and an exposure apparatus using the method wherein high accuracy alignment is provided by switching measurement conditions or measurement parts for the alignment in measuring alignment marks formed in each layer. In order to achieve the object above-described, an alignment method of the present invention is a method wherein when forming a new layer on a substrate, alignment is performed by measuring each position of existing layers formed prior to the above-described new layer and the above-described new layer in a first measurement condition or a second measurement condition, comprising the steps of: measuring by switching between the above-described first and second measurement conditions for marks formed in each of the above-described existing layers; and performing alignment between the above-described existing layers and the above-described new layer based on measurement of mark positions of the above-described existing layers. Preferably, the above-described second measurement condition has a plurality of different conditions in an optical characteristic, and the measurement is performing by switching the measurement conditions. As the optical characteristic, preferably wavelength of illumination light for the measurement is switched. As the above-described optical characteristic values representing light intensity distribution of illumination light for the measurement (σ=standard deviation) may be switched. The present invention includes an exposure apparatus for using any of the above-described alignment method and forming the above-described new layer. The exposure apparatus according to the present invention is an apparatus wherein an exposed object is aligned based on measurement of position information on marks formed in each of existing layers on the exposed object on which the existing layers are provided and a new layer is to be formed, and then projection exposure is performed, the apparatus having a first measurement part and a second measurement part for measuring the position information on the marks, the above-described first and second measurement parts being configured such that they can be switched for the marks formed in each of the above-described existing layers. It is preferred that the above-described first and second measurement parts are switched manually, or that switching of the above-described first and second measurement parts are performed based on automatic calculation of contrast that is made before exposure. An exposure method of the present invention is a method for aligning an exposed object having a plurality of existing layers with alignment marks formed in each of them based on measurement of the alignment marks, and projection-exposing the object, wherein when measuring the alignment marks in the above-described each layer, each alignment mark is measured by switching conditions of illumination light for the measurement depending on the particular alignment mark in each layer. The present invention can also be applied to a semiconductor device manufacturing method comprising the steps of: installing in a semiconductor manufacturing factory a plurality of manufacturing apparatuses for performing various processes including the above-described exposure apparatus; and manufacturing semiconductor devices by performing a plurality of processes with the manufacturing apparatuses. The semiconductor device manufacturing method may be also characterized in that it further comprises the steps of: connecting the above-described manufacturing apparatuses by a local area network; and data-communicating information about at least the above-described manufacturing apparatuses of the above-described manufacturing apparatuses between the above-described local area network and an external network outside the above-described semiconductor manufacturing factory, characterized in that maintenance information for the above-described exposure apparatus is obtained by accessing and communicating data with a database provided by a vendor or user of the above-described exposure apparatus via the above-described external network, or production control is conducted by communicating data via the above-described external network between the above-described semiconductor manufacturing factory and a semiconductor manufacturing factory other than the above-described semiconductor manufacturing factory. The present invention may be applied to a semiconductor manufacturing factory comprising: a group of manufacturing apparatuses for performing various processes including the above-described exposure apparatuses; a local area network for connecting the manufacturing apparatuses; and a gateway allowing access by the local area network to an external network outside of the factory, wherein data communication of information about at least one apparatus of the above-described manufacturing apparatuses is provided, and the present invention can be applied to a maintenance method for an exposure apparatus installed in a semiconductor manufacturing factory, comprising the steps of: a user or vendor of the above-described exposure apparatus providing a maintenance database connected to an external network for the semiconductor manufacturing factory; allowing access from inside of the above-described semiconductor manufacturing factory via the above-described external network to the above-described maintenance database; and transmitting maintenance information stored in the above-described maintenance database via the above-described external network to the semiconductor manufacturing factory side. The present invention may also be characterized in that the above-described exposure apparatus further comprises a display, a network interface, and a computer for executing software for the network, wherein data-communication of maintenance information on the exposure apparatus via a computer network is provided, and preferably, the software for the network provides on the above-described display a user interface connected to an external network for a factory with the above-described exposure apparatus installed therein, for accessing a maintenance database provided by a vendor or user of the above-described exposure apparatus, whereby allowing acquisition of information from the database via the above-described external network. Other features and advantages of the present invention will be apparent from the following description taken in conjunction with the accompanying drawings, in which like reference characters designate the same or similar parts throughout the figures thereof. Other objects and advantages besides those discussed above shall be apparent to those skilled in the art from the description of a preferred embodiment of the invention which follows. In the description, reference is made to the accompanying drawings, which form a part thereof, and which illustrate an example of the invention. Such an example, however, is not exhaustive of the various embodiments of the invention, and, therefore, reference is made to the claims which follow the description for determining the scope of the invention. An alignment method according to an embodiment of the present invention switches measurement conditions or measurement parts in measuring alignment marks formed in each layer. In the past, in exposure apparatuses, as a measurement mechanism for alignment marks to obtain information about position in a wafer surface, there are known an Off-Axis mechanism wherein non-exposure light is used and does not pass any projection lens, and a non-exposure TTL (Through The Lens) mechanism wherein non-exposure light is used and passes a projection lens. Particularly, multiple wavelengths are recently used as illumination light for observation in order to improve detection accuracy for AA mark observation images. For example, the use of a light source with a relatively wide wavelength width by a halogen lamp (633±30 nm, for example) as illumination light for observing non-exposure light is characterized in that an interference fringe by a resist film, which tends to occur when observing an AA mark on a wafer by using a monochromatic light source such as a He—Ne laser, can be reduced, thereby providing better alignment. There is also a method to improve the interference conditions by varying the center wavelength of illumination light. It is also possible to vary illumination conditions for a He—Ne laser, for example, conditions for a value representing a light intensity distribution of illumination (σ=standard deviation). Thus, in order to support processes that vary for each layer, each alignment mark is measured by switching conditions of illumination light optimized for alignment marks in each layer when measuring alignment marks in each layer. Various embodiments will be now described with reference to the drawings. FIG. 1 shows an exposure apparatus comprising an alignment device including a non-exposure TTL mechanism and an Off-Axis mechanism, according to a first embodiment of the present invention. As shown in FIG. 1, light from a light source emitted by illumination system 1 irradiates a reticle 2 as a first object. Patterns on a surface of the reticle 2 are projection-transferred onto wafer 4 as a second object by a projection optical system 3. The wafer 4 is fixed onto a movable stage 5 disposed on a base plate 10. The reticle 2 is held on a reticle stage 2a, and the reticle stage 2a is moved in X, Y, Z, and θ directions by means of a reticle drive mechanism (not shown). On the wafer 4, a plurality of alignment marks 4a and 4b have been formed in existing layers until the last process step, which marks are located in a certain positional relationship with respect to circuit patterns. Reference numeral 6 denotes a microscope with a TTL mechanism as a first measurement condition or measurement part. The TTL microscope 6 is a microscope that reads the alignment mark 4a on the wafer 4 through lenses of the projection optical system 3, and performs position measurement for the alignment mark 4a on the wafer 4 by a mirror 6d located between the reticle 2 and the projection optical system 3. Reference numeral 6a denotes an illumination light source with a He—Ne laser, and reference numeral 6b is a CCD camera. The illumination light emitted by the illumination light source 6a is transmitted by an alignment optical system 6c and mirror 6d disposed between the reticle 2 and projection optical system 3, through lenses in the projection optical system 3, thereby illumination alignment mark regions on the wafer 4. Light reflected or scattered on the wafer 4 based on the alignment mark 4a is transmitted again through the lenses in the projection optical system 3, by the mirror 6d and alignment optical system 6c, and then is imaged on the CCD camera 6b. The CCD camera 6b processes the image of the alignment mark 4a to obtain position information on the wafer 4. Reference numeral 7 denotes an Off-Axis microscope as a second measurement condition or measurement part, which microscope has two measurement methods. It is placed at a position other than the TTL type microscope 6 and measures an alignment mark 4b formed at a position in a certain positional relationship with respect to an exposed position. Reference numeral 7a denotes a first illumination light source with a He—Ne laser used in a first measurement method, and reference numeral 7b denotes an illumination light source with a halogen lamp used in a second measurement method, and reference numeral 7c denotes a CCD camera. One of these two light sources 7a and 7b used in the first and second measurement methods is selected as a light source, and illumination light from the light source illuminates an alignment mark region on the wafer 4 through an alignment optical system 7d without passing through the projection optical system 3. The light reflected or scattered on the surface of the wafer 4 is again imaged through the alignment optical system 7d on the CCD camera 7c. The position information is measured by image-processing with the CCD camera 7c. FIGS. 2A and 2B show examples of arrangements of alignment marks of this embodiment. FIG. 2A shows an in-shot plan arrangement, and FIG. 2B shows a step formation of the alignment marks. Reference numerals 11 and 13 denote alignment marks formed in layer A, and reference numerals 12 and 14 denote alignment marks formed in layer B. FIG. 3 shows an example of an arrangement for measurement shots S1 to S8 in global alignment, and FIG. 4 shows a measurement/exposure sequence in the example. After selecting a reticle set at step 31, a wafer set at step 32, and a halogen lamp as the light source at step 33, at first using the Off-Axis microscope 7, while halogen lamp 7b with a wide wavelength band is being set as the light source, alignment mark 11 formed in layer A shown in FIGS. 2A and 2B is measured at step 34, and alignment mark 13 formed in layer A shown in FIGS. 2A and 2B is measured at step 35. Next, after selecting He—Ne laser light as the light source at step 36, while the He—Ne laser light 7a is being set as the light source, the alignment mark 12 formed in layer B is measured at step 37 and the alignment mark 14 is measured at step 38. A determination is made of all sample shots having been measured at step 39, so that the measurements described above are repeated until measurement of all sample shots is completed. Upon completion of measurement of all sample shots, deviations of the reticle with respect to layer A is calculated from measurements of the mark 11 and mark 13 at step 40, and deviations of the reticle with respect to layer B is calculated from measurements of the mark 12 and the mark 14 at step 41. After statistically processing the amount of the deviations at step 42, and correction-driving the reticle stage at step 43, exposure is performed at step 44, and the wafer is unloaded at step 45, and a determination is made if the process is completed for the whole wafer at step 46, and if the process is not completed for the whole wafer, then the processes described above are repeated starting with the wafer setting at step 32. For example, measurement conditions or measurement parts are predetermined such that for each mark, switching of light sources such as a He—Ne laser and a halogen lamp, as the measurement conditions or measurement parts, provides respective waveforms of the marks detected by the CCD camera with good contrast, and a manual switching part is provided such that the measurement conditions or measurement parts can be set in the apparatus side. In a second embodiment of the present invention, a TTL mechanism and an Off-Axis mechanism can be switched as measurement conditions or measurement parts, and a wavelength filter for varying the center wavelength of a halogen lamp may be formed as a measurement method. It is also possible to switch values representing a light intensity distribution condition of illumination by He—Ne laser light (σ=standard deviation, not shown) and to change them. As in the first embodiment, measurement conditions or measurement parts are predetermined such that for each mark, switching of the TTL mechanism and Off-Axis mechanism, switching of center wavelengths of the halogen lamp, or switching of as, as the measurement conditions or measurement parts, provides waveforms for the marks with good contrast, and a manual switching part is provided such that the measurement conditions or measurement parts can be set in the apparatus side. It is also possible to automatically calculate the contrast and determine for each mark an optimized measurement condition or measurement part, or a wavelength and σ. In measurement of a wafer position, it is needed to set an optimized illumination condition since reflection, absorption, scattering, diffraction and interference of the illumination light affect the measurement of the wafer position depending on the physical features of the alignment marks and manufacturing process such as application condition of a resist. In a third embodiment of the present invention, it is possible to automatically switch measurement conditions for each of the marks formed in the wafer right after starting each lot processing and to automatically calculate contrast to determine an optimized measurement condition or measurement part for each mark. As shown in FIG. 6A, the mark is composed of four rectangular portions having the same geometry. As described for the first embodiment, light flux reflected by the alignment mark transmits the lenses of the projection optical system 3 through the alignment optical system, and then forms an alignment mark image WM on the CCD camera. It is subjected to photo-electric conversion in the CCD camera, then converted to a two-dimensional digital sequence in an A/D conversion device (not shown). Then, a process window Wp is set for the digital signal conversion, as shown in FIGS. 6A and 6B, and the two-dimensional image signal is converted to a one-dimensional digital signal sequence S(x) by an addition process in the y-direction. The contrast is changed for each mark, as shown in FIGS. 7A and 7B, by switching the alignment measurement conditions, for example, by switching light sources such as a halogen lamp or He—Ne laser, or by switching light intensity distribution conditions σ for He—Ne laser light. FIG. 7A shows an image signal with high contrast while FIG. 7B shows an image signal with low contrast. FIG. 5 is a flowchart showing a process procedure for determining an optimized measurement condition for each mark. As shown in FIG. 5, a halogen lamp is selected as the light source at step 502, and measurement of an alignment mark is performed with the Off-Axis microscope 7 at step 503, and a contrast value is calculated at step 504. The measurement and contrast calculation are repeated for the alignment marks 11 to 14 at step 505. Next, the center wavelength of the halogen lamp is varied at step 506, and then a similar measurement of the alignment marks are performed at step 507. A similar contrast calculation is performed at step 508. The measurement and contrast calculation are repeated for the alignment marks 11 to 14 at step 509. Steps 510 to 513 are process steps in the case where He—Ne laser light is selected as the alignment light source. After the He—Ne laser is selected at step 510, a similar procedure is performed by steps 511 to 513, as performed by steps 503 to 505 described above. Steps 514 to 517 are process steps in the case where a light intensity distribution a of the He—Ne laser light is varied. After the light intensity distribution σ is varied at step 514, a similar procedure is performed by steps 515 to 517, as performed by steps 503 to 505 described above. At step 517, an optimized measurement condition is determined for each mark from contrast values for varied measurement conditions. The determined measurement condition is held in the lot, and global alignment is conducted with the determined conditions, for each wafer. The measurement may be automatically performed for each lot with the determined measurement conditions, or the measurement condition may be held as recipe conditions for each lot. In the case of starting the same lot, the time for the automatic measurement is reduced by referring to the held measurement conditions. An example of a production system for semiconductor devices (e.g., semiconductor chips such as ICs and LSIs, liquid crystal panels, CCDs, thin film magnetic heads, micro-machines, etc.) by an apparatus according to the present invention will be described. In the system, maintenance services such as trouble management or regular maintenance, or provision of software for manufacturing apparatuses installed in semiconductor manufacturing factories are provided by using a computer network outside the manufacturing factories. FIG. 8 is a representation picked up with a certain angle from the whole system. In the figure, reference numeral 101 denotes an office of a vendor (e.g., an apparatus supplier/manufacturer) providing manufacturing apparatuses for manufacturing semiconductor devices. It is assumed that the example of manufacturing apparatuses includes semiconductor manufacturing apparatuses for performing various processes used in a semiconductor manufacturing factory, for example, apparatuses for pre-process (e.g., lithography apparatuses such as exposure apparatuses, resist process apparatuses, and etching apparatuses, heat treatment apparatuses, film deposition apparatuses, planarization apparatuses, etc.), and apparatuses for post-process (e.g., assembly apparatuses, inspection apparatuses, etc.). The office 101 comprises a host management system 108 for providing a maintenance database for manufacturing apparatuses, a plurality of operation terminal computers 110, and a local area network (LAN) 109 to construct an intranet or the like. The host management system 108 comprises a gateway for connecting the LAN 109 to Internet 105 (external network of the office) and a security function to limit access from the outside. Reference numerals 102 to 104 denote manufacturing factories of semiconductor manufacturers as users of the manufacturing apparatuses. The manufacturing factories 102 to 104 may be factories belonging to different manufacturers, or may also be factories belonging to a single manufacturer (for example, a factory for pre-process and a factory for post-process). Each of the factories 102 to 103 comprises a plurality of manufacturing apparatuses 106, a local area network (LAN) 111 for connecting them to construct an intranet or the like, and a host management system 107 as a monitor apparatus for monitoring operation status of each manufacturing apparatus 106. The host management system 107 provided in each of the factories 102 to 104 comprises a gateway for connecting LAN 111 in each factory to Internet 105 (external network of the factories). This allows access from the LAN 111 via Internet 105 to the host management system 108 in the vendor 101's side, and the security function in the host management system 108 allows only a predefined user to access it. Specifically, notification from the factory to the vendor, of status information representing operation status of each manufacturing apparatus 106 (for example, symptoms of the manufacturing apparatus with trouble occurrence), as well as reception from the vendor of response information responding to the notification (for example, information indicating management methods for trouble, software and data for the management for the trouble) and maintenance information such as up-to-date software and help information, are possible. For data communication between each of factories 102 to 104 and vendor 101, data communication over LAN 111 in each factory, a communication protocol (TCP/IP) commonly used in the Internet is employed. Instead of utilizing the Internet as an external network outside the factory, a dedicated line network (such as an ISDN), which tightens security to avoid access by a third party, may also be utilized. The host management system is not limited to the one provided by the vendor. The user may also construct a database and place it on an external network, and to allow access from a plurality of factories of the user to the database. FIG. 9 shows a concept representation picked up with a different angle than FIG. 8 from the whole system of this embodiment. In the example described above, the plurality of user factories each having manufacturing apparatuses and the management system of the vendor of the manufacturing apparatuses are connected to each other via the external network, and information about production control of each factory and at least one manufacturing apparatus is data-communicated via the external network. On the other hand, in this example, a factory comprising apparatuses from a plurality of vendors and a management system of the vendor of each of the plurality of manufacturing apparatuses are connected to each other via an external network outside the factory, and maintenance information for each manufacturing apparatus is data-communicated. In the figure, reference numeral 201 denotes a manufacturing factory of a user of manufacturing apparatuses (e.g., a semiconductor device manufacturer), a manufacturing line of which is provided with manufacturing apparatuses for performing various processes, for example, here, exposure apparatuses 202, resist process apparatuses 203, and a film deposition process apparatus 204. While FIG. 9 shows only one manufacturing factory 201, actually, multiple factories are connected by a network as well. Each apparatus in the factory is connected to the others via LAN 206 to constitute an intranet, and operation management of the manufacturing line is conducted by a host management system 205. On the other hand, offices of vendors (e.g., apparatus supplier/manufacturers) such as an exposure apparatus manufacturer 210, a resist process apparatus manufacturer 220, and a film deposition apparatus manufacturer 230, have host management systems 211, 221, and 231, respectively, for conducting remote maintenance for the apparatuses supplied by respective manufacturers, and the host management systems comprise a respective maintenance database and a gateway to an external network, as described above. The host management system 205 for managing each apparatus in the user's manufacturing factory is connected via the Internet or a dedicated line network (external network 200) to management systems of a vendor of apparatuses 211, 221, and 231, respectively. In this system, although upon trouble occurrence in any of the group of apparatuses in the manufacturing line, the operation of the manufacturing line stops, a quick measure can be implemented by receiving remote maintenance via the Internet 200 from the vendor of the apparatus with the trouble occurrence, thereby minimizing the stoppage of the manufacturing line. Each of the manufacturing apparatuses installed in the semiconductor manufacturing factory has a display, a network interface, and a computer for executing network access software and apparatus operation software stored in a storage device. The storage device is such as a built-in memory, hard-disc, or network file server. The above-described network access software includes web browsers for dedicated or general purposes, and provides on its display a user interface with a picture such as one illustrated in FIG. 10, for example. An operator who manages the manufacturing apparatuses in each factory, referring to the picture, inputs information such as machine-type of apparatus 401, serial number 402, title of trouble 403, day of occurrence 404, degree of emergency 405, symptom 406, measure 407, history 408, and so on, into input items on the picture. The input information is transmitted via the Internet to the maintenance database, then suitable maintenance information of the result is sent back from the maintenance database to be presented on the display. The user interface provided by the web browser also implements hyperlink functions 410 to 412, as shown in the figure, thereby allowing the operator to access more detailed information on each item, to retrieve up-to-date version software to be used for the manufacturing apparatus from a software library provided by the vendor, or to retrieve an operation guide (help information), which is provided for factory operators to refer to. Here, the maintenance information provided by the maintenance database also includes information related to the present invention described above, that is to say, information about suitable measurement conditions for each preformed layer and information about the measurement parts or measurement conditions, and the above-described software library also provides up-to-date software for implementing the present invention. A manufacturing process for manufacturing semiconductor devices utilizing the production system mentioned above will be described. FIG. 11 shows the overall flow of a manufacturing process for the semiconductor devices. Circuit design for a semiconductor device is conducted at step 1 (circuit design). Masks with the designed circuit pattern formed thereon are fabricated at step 2 (mask fabrication). On the other hand, wafers are prepared with a material such as silicon at step 3 (wafer preparation). Step 4 (wafer process) is referred to as a pre-process, and actual circuits are formed on the wafers by lithography techniques with the prepared masks described above and the wafers. The next step, step 5 (assembly), is referred to as a post-process, which is a process to form semiconductor chips by using the wafers fabricated by step 4, and includes processes for assembly such as assembly processes (dicing, bonding), and a packaging process (chip encapsulation). At step 6 (inspection), inspections such as a performance verification test or a durability test for the semiconductor devices fabricated at step 5 are carried out. The semiconductor devices, through those processes, are completed, and then shipped (step 7). The pre-process and post-process are respectively conducted in respective different dedicated factories, and maintenance is performed for each factory by the remote maintenance system described above. Information for production control or apparatus maintenance is also data communicated through the Internet or a dedicated line network between the pre-process factory and the post-process factory. FIG. 12 shows a detailed flow of the above-described wafer process. At step 11 (oxidation), a surface of a wafer is oxidized. At step 12 (CVD), an insulator film is deposited on the wafer surface. At step 13 (electrode formation), electrodes are formed on the wafer surface by evaporation. At step 14 (ion implantation), ions are implanted into the wafer. At step 15 (resist process), a photosensitive agent is applied onto the wafer. At step 16 (exposure), the wafer is printed-exposed with the circuit pattern on the mask by the above-described exposure apparatus. At step 17 (development), the exposed wafer is developed. At step 18 (etching), portions other than the developed resist pattern are etched off. At step 19 (resist strip), a resist after etching, which is no longer necessary, is removed. Multiple circuit patterns are formed on the wafer by repeating those steps. Since the manufacturing apparatuses used for each process are maintained by the above-described remote maintenance system, trouble can be prevented beforehand, and when trouble occurs, performance can quickly be recovered, thereby improving productivity for semiconductor devices as compared with conventional ways. The present invention is not limited to the above embodiments and various changes and modifications can be made within the spirit and scope of the present invention. Therefore, to apprise the public of the scope of the present invention the following claims are made. |
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046506373 | claims | 1. A method of locating a defective fuel rod within a nuclear fuel assembly that is leaking radioactive products into surrounding coolant of a liquid coolant bath in which said fuel assembly is immersed, wherein said fuel rod is one of a plurality of fuel rods interconnected in a generally parallel spaced-apart matrix in the fuel assembly, said method comprising the steps of: (a) physically isolating successive longitudinal segments of said matrix of fuel rods from adjacent portions of said coolant bath at least along two opposing lateral sides of said matrix; (b) selectively sampling the coolant surrounding said plurality of fuel rods within each of said successive physically isolated longitudinal segments of said matrix to determine the approximate one of said successive longitudinal segments at which radioactive products are being emitted from a fuel rod therein; (c) physically isolating successive submatrices of fuel rods at least along two opposing lateral sides of each of said submatrices from adjacent fuel rods in said approximate one of said successive longitudinal segments of said matrix at which it was determined that radioactive products were being emitted from a fuel rod therein; and (d) selectively sampling the coolant surrounding said fuel rods within each of said successive physically isolated submatrices to determine the one of said successive submatrices at which radioactive products are being emitted from a fuel rod therein. (a) drawing a sample of said coolant from a volumetric test zone defined respectively by each of said physically isolated longitudinal segments of said matrix of fuel rods and each of said physically isolated submatrices of said approximate one of said longitudinal segments of said matrix of fuel rods; and (b) measuring the radioactive product content of said drawn sample. (a) placing a coolant sample probe of the type having a pair of baffles extending in cantilevered manner from a collector member to free ends and defining therebetween a volumetric test zone, in cooperative proximity to said fuel assembly rods and intermediate said first and second ends thereof such that a longitudinal segment of at least one of said fuel rods is disposed within said volumetric test zone and physically isolated by said baffles from any fuel rods in said assembly being located outside of said test zone; and (b) sampling the radioactive content of liquid coolant of said volumetric test zone when said sample probe is positioned as in step (a). (a) a pair of opposed baffles longitudinally extending between first and second ends; (b) a collector operatively connecting said baffles adjacent said first ends thereof so as to define a volumetric test zone between said opposed baffles; said collector being configured to collect liquid coolant drawn from said volumetric test zone; and (c) wherein said probe apparatus is configured for operative alignment with a nuclear fuel rod assembly such that said baffles laterally project on opposite sides of at least one fuel rod of said assembly so as to encompass a longitudinal segment of said fuel rod within said volumetric test zone and physically isolate said longitudinal segment of said fuel rod within said test zone from any fuel rods in said assembly being located outside of said test zone. (a) determining the approximate one of a succession of longitudinal segments of said matrix of fuel rods at which radioactive products are being emitted from a fuel rod of said matrix; (b) physically isolating successive submatrices of fuel rods at least along two opposing lateral sides of each of said submatrices from adjacent fuel rods in said approximate one longitudinal segment of said matrix at which it was determined that radioactive products were being emitted from a fuel rod therein; and (c) selectively sampling the coolant surrounding said fuel rods within each of said successive physically isolated submatrices to determine the one of said successive submatrices at which radioactive products are being emitted from a fuel rod therein. (a) drawing a sample of said coolant from said volumetric test zone defined respectively by each of said physically isolated submatrices of said matrix of fuel rods; and (b) measuring the radioactive product content of said drawn sample. 2. The method as recited in claim 1, wherein the approximate one of the longitudinal segments of said matrix of fuel rods in said fuel assembly containing the leaking fuel rod is the longitudinal segment which contains the coolant with the highest sampled radioactive product content. 3. The method as recited in claim 1, wherein said step of physically isolating successive longitudinal segments of said matrix of fuel rods comprises defining a volumetric test zone about each one of said successive longitudinal segments wherein said volumetric test zone in one direction circumferentially encompasses the fuel rods in said matrix thereof in a plane transverse to the longitudinal axes of the fuel rods and in another direction longitudinally extends in the axial direction of said fuel rods a distance substantially less than the length of said fuel rods. 4. The method as recited in claim 1, wherein said each sampling step comprises: 5. The method as recited in claim 1, wherein the sampling for said radioactive products of coolant surrounding said fuel rods within said submatrices is carried out in logical manner so as to isolate that fuel rod that is emitting said radioactive products into its surrounding coolant. 6. The method as recited in claim 1, wherein said step of physically isolating successive submatrices of fuel rods comprises selectively inserting one or more baffle members between adjacent rows of said fuel rods within said matrix thereof so as to isolate the coolant surrounding said fuel rods within one submatrix from that of an adjacent submatrix. 7. The method as recited in claim 6, wherein said baffle members are inserted between said adjacent fuel rod rows in a direction transverse to the longitudinal axes of said fuel rods. 8. A method of locating a leaking fuel rod in a nuclear fuel rod assembly of the type having a plurality of elongate fuel rods longitudinally extending between first and second ends in generally parallel spaced-apart manner and disposed within a liquid coolant bath, comprising the steps of: 9. The method as recited in claim 8, wherein said fuel rod assembly has a cross-sectional area defining a matrix of a plurality of said fuel rods, and wherein said coolant sample probe is sized and configured and is placed in cooperative proximity with said fuel rod assembly such that longitudinal segments of substantially the entire matrix of said plurality of fuel rods are disposed within said volumetric test zone. 10. The method as recited in claim 9, wherein said baffles are characterized by an effective length dimension extending from said collector member toward said free ends and an effective width dimension generally orthogonally disposed to said length dimension and substantially less than the length of said fuel rods as measured between said first and said second ends thereof; wherein said baffles are disposed such that their width dimension lies generally parallel with the longitudinal axes of said fuel rods when said coolant sample probe is operatively positioned in close proximity to said fuel assembly rods. 11. The method as recited in claim 10, including the step of longitudinally moving said coolant sample probe relative to said fuel rod assembly such that said volumetric test zone traverses and said sampling is performed over substantially the entire length of said fuel rods between said first and said second ends thereof. 12. The method as recited in claim 11, including the step of determining that longitudinal position of said coolant sample probe relative to said fuel assembly where said radioactive content of said liquid coolant sample from said volumetric test zone is a maximum. 13. The method as recited in claim 8, wherein the step of placing said coolant sample probe in cooperative proximity to said fuel assembly rods comprises the step of moving said probe member relative to said fuel rod assembly such that at least one of said baffle members cooperatively slides between adjacent rows of said fuel rods, thereby isolating the longitudinal segments of those fuel rods within said volumetric test zone from those fuel rods outside of said test zone. 14. The method as recited in claim 13, including the step of systematically moving said test probe relative to said fuel rod assembly so as to isolate longitudinal segments of different groups of said fuel rods, and performing said sampling step on each of said groups of said fuel rods. 15. The method as recited in claim 14, wherein said fuel rod assembly has a cross-sectional area defining a matrix of a plurality of said fuel rods, and wherein said coolant sample probe is sized and configured to encompass approximately one-half of said fuel rods of said matrix within said volumetric test zone. 16. The method as recited in claim 15, further including the step of systematically moving said test probe and performing said sampling so as to subdivide said matrix of fuel rods into test quadrants. 17. The method as recited in claim 8, wherein the step of sampling includes drawing a sample of liquid coolant from said volumetric test zone and testing the radioactive content of said drawn coolant sample. 18. Test probe apparatus for determining the position of leaking fuel rods in a nuclear fuel rod assembly of the type having a plurality of elongate spaced fuel rods interconnected in generally parallel configuration, comprising: 19. Test probe apparatus as recited in claim 18, further including means cooperatively connected with said collector for drawing liquid coolant from said volumetric test zone through said collector. 20. Test probe apparatus as recited in claim 19, further including sensor means operatively connected with said collector for measuring the radioactive content of said liquid coolant drawn through said collector. 21. Test probe apparatus as recited in claim 18, further including means operatively connected with said baffles and collector for moving said volumetric test zone longitudinally relative to said fuel assembly rods. 22. Test probe apparatus as recited in claim 18, further including means operatively connected with said baffles and collector for moving said volumetric test zone laterally relative to said fuel assembly, whereby the number of said longitudinal segments of said fuel rods contained within said volumetric test zone at any instant of time can be selectively varied. 23. Test probe apparatus as recited in claim 22, further including means operatively connected with said baffles and collector for moving said volumetric test zone longitudinally relative to said fuel assembly rods. 24. Test probe apparatus as recited in claim 18, wherein said baffles are constructed of thin, semirigid sheet material having a thickness less than 0.20 inches. 25. Test probe apparatus as recited in claim 18, wherein said baffles are generally planar, and are disposed parallel to one another, and wherein at least one of said baffles is constructed of thin sheet material having a thickness sized to cooperatively slide between adjacent ones of said fuel rods within said fuel assembly. 26. Test probe apparatus as recited in claim 25, wherein said baffles are constructed of thin, semirigid sheet material having a thickness less than 0.20 inches. 27. Test probe apparatus as recited in claim 18, wherein said baffles are constructed of thin, semirigid sheet material having a thickness less than 0.10 inches. 28. Test probe apparatus as recited in claim 18, wherein the fuel assembly with which the probe is to be used has a cross-sectional area dimension as measured in a plane generally perpendicular to the longitudinal axes of said fuel rods and a longitudinal length dimension generally equal to that of the plurality of fuel rods comprising said fuel assembly; wherein said baffles are disposed generally parallel to one another; wherein said volumetric test zone has a cross-sectional area dimension as measured in a plane perpendicular to said baffles and extending through said collector which is larger than said fuel assembly cross-sectional area; and wherein the height of said baffles as measured perpendicular to the volumetric test zone cross-sectional dimension is significantly less than the longitudinal length of said fuel assembly. 29. Test probe apparatus as recited in claim 18, wherein the fuel assembly with which the probe is to be used has a cross-sectional area dimension as measured in a plane generally perpendicular to the longitudinal axes of said fuel rods and a longitudinal length dimension generally equal to that of the plurality of fuel rods comprising said fuel assembly; wherein said baffles are disposed generally parallel to one another; wherein said volumetric test zone has a cross-sectional area dimension as measured in a plane perpendicular to said baffles and extending through said collector which is approximately one-half that of said fuel assembly cross-sectional area; and wherein the height of said baffles as measured perpendicular to the volumetric test zone cross-sectinal dimension is significantly less than the longitudinal length of said fuel assembly. 30. A method of locating a defective fuel rod within a nuclear fuel assembly that is leaking radioactive products into surrounding coolant of a liquid coolant bath in which said fuel assembly is immersed, wherein said fuel rod is one of a plurality of fuel rods interconnected in a generally parallel spaced-apart matrix in the fuel assembly, said method comprising the steps of: 31. The method as recited in claim 30 wherein said step of physically isolating successive submatrices of said approximate one longitudinal segment of said matrix of fuel rods comprises defining a volumetric test zone about each one of said successive submatrices wherein said volumetric test zone in one direction circumferentially encompasses the fuel rods in said submatrix thereof in a plane transverse to the longitudinal axes of the fuel rods and in another direction longitudinally extends in the axial direction of said fuel rods a distance substantially less than the length of said fuel rods. 32. The method as recited in claim 31, wherein said each sampling step comprises: 33. The method as recited in claim 30, wherein the sampling for said radioactive products of coolant surrounding said fuel rods within said submatrices is carried out in logical manner so as to isolate that fuel rod that is emitting said radioactive products into its surrounding coolant. 34. The method as recited in claim 30, wherein said step of physically isolating successive submatrices of fuel rods comprises selectively inserting one or more baffle members between adjacent rows of said fuel rods within said matrix thereof so as to isolate the coolant surrounding said fuel rods within one submatrix from that of an adjacent submatrix. |
050930753 | abstract | Upper internals for a pressurized nuclear reactor include an assembly for collecting the coolant leaving the core and a separating device. The device has guides for the control clusters and their drive shafts, a lower plate formed with passages for the coolant leaving the core and an upper plate formed with holes for passage of the same coolant into the collection assembly or plenum. A peripheral shroud may connect the plates together. Each cluster guide of the flow separation device has an internal casing devoid of openings, having a practically coolant-tight connection with the plates, whereby the control clusters and their drive shafts are protected against the high speed coolant flow from the core to the outlet nozzle. |
description | This application is a continuation-in-part of copending U.S. patent application Ser. No. 14/038,424 entitled “Recovering and Recycling Uranium Used for Production of Molybdenum-99,” filed Sep. 26, 2013, incorporated by reference herein. This invention was made with government support under Contract No. DE-AC52-06NA25396 awarded by the U.S. Department of Energy. The government has certain rights in the invention. The present invention relates generally to the recovery of uranium from an irradiated solid target and more particularly to the recovery and purification of uranium from an irradiated solid target after removal of molybdenum-99 produced from the target. Technetium-99m (“Tc-99m”) is the most commonly used radioisotope in nuclear medicine. Tc-99m is used in approximately two-thirds of all imaging procedures performed in the United States. Tens of millions of diagnostic procedures using Tc-99m are undertaken annually. Tc-99m is a daughter isotope produced from the radioactive decay of molybdenum-99 (“Mo-99”). Mo-99 decays to Tc-99m with a half life of 66 hours. The vast majority of Mo-99 used in nuclear medicine in the U.S. is produced in aging foreign reactors. Many of these reactors still use solid highly enriched uranium (“HEU”) targets to produce the Mo-99. HEU has a concentration of uranium-235 (“U-235”) of greater than 20%. Maintenance and repair shutdowns of these reactors have disrupted the supply of Mo-99 to the U.S. and to most of the rest of the world. The relatively short half-life of the parent radioisotope Mo-99 prohibits the build-up of reserves. One of the major producers, The National Research Reactor in Canada, will cease production in 2016. An alternative strategy for providing Mo-99 is based upon the use of low enriched uranium (LEU), which presents a much lower nuclear proliferation risk than HEU. LEU has a concentration of U-235 of less than 20%, and many international Mo-99 producers are converting from HEU to LEU solid targets for Mo-99 production. Several of the technologies currently being considered for the domestic supply of Mo-99 are based on the fission of U-235 in LEU. In all processes being considered, only a small fraction of the U-235 present in the irradiated target will be consumed during irradiation. Fission of U-235 generates a variety of fission products, one of which is Mo-99. Some form of enriched uranium (HEU and/or LEU) is used for the production of Mo-99. After the fission process, the remaining uranium is typically discarded along with other fission products as waste. Recovery and purification of the uranium would make it available for reuse, storage, or disposal. Therefore, an object of the present invention is to provide a process for recovering, and purifying, uranium from an irradiated solid target after separating Mo-99 produced from the irradiated target. The embodiments for recovering uranium apply to recovering all isotopic ratios of uranium, including low-enriched uranium (LEU) as well as highly-enriched uranium (HEU). Enriched uranium refers to uranium enriched in isotope U-235. An embodiment relates to a process for recovering uranium from an irradiated solid target, after recovering Mo-99 produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising Mo-99, and thereafter dissolving the target. Following dissolution, the solution is conditioned to provide an aqueous nitric acidic solution comprising a first acid concentration and a first uranium concentration. The uranium in the acidic solution will be in the +VI oxidation state and in the chemical form of the uranyl di-oxo di-cation (UO22+). The acidic solution, along with the uranium, will pass through a solid sorbent, while Mo-99 is removed from the solution, remaining adsorbed to the sorbent. The Mo-99 will be recovered in a subsequent desorption step. After passing through the sorbent, the concentration of acid and uranium in the acidic uranium solution is adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing this crystallization of uranyl nitrate hydrates, the uranium contained in the uranyl nitrate hydrates is separated from a supernatant which contains soluble fission products. Thus the uranium is recovered and purified during this process, and is available for reuse, storage, or disposal. The embodiment process for recovering uranium applies to recovering all isotopic ratios of uranium including LEU as well as HEU. Enriched uranium refers to uranium enriched in isotope U-235. An embodiment process relates to recovery of uranium that has been used for the production of Mo-99 generated from the fission of U-235. Mo-99 undergoes radioactive decay to Tc-99m, the most widely used radioisotope in nuclear medicine. Recovery and purification of uranium allows for its reuse, storage, or disposal. It should be understood that uranium includes both LEU (uranium having less than 20% of the U-235 isotope), and also HEU (uranium having greater than 20% of the U-235 isotope). Thus, an embodiment of the disclosed process may be used for recovery of either LEU or HEU. An embodiment relates to a process for recovering uranium from a solid target that has been used for the production of Mo-99. The process employs a sorbent-based separation. The sorbent is used to remove Mo-99 prior to recovery and purification of the uranium. The process begins with irradiation of a solid target having fissionable uranium (i.e., U-235). The irradiation promotes fission of the U-235 to form fission products that include Mo-99. After the irradiation the solid target is dissolved. The resultant solution is conditioned to provide an aqueous nitric acid solution of from about 0.01 M to about 2 M (M means moles of nitric acid per liter of solution). The uranium concentration of this solution is from about 50 gU/L to about 350 gU/L (gU/L means grams of uranium per liter of solution). The acidic solution, along with the uranium, will pass through a solid sorbent (e.g., a titania-based sorbent or an alumina-based sorbent), while Mo-99 is removed from the solution, remaining adsorbed to the sorbent. The Mo-99 will be recovered in a subsequent desorption step (e.g., washing the sorbent with an alkaline solution to strip the Mo-99 from the sorbent). The sorbent may be packed into a column, with processing solutions then flowing through the column. After passing through the sorbent, and removal of the majority of the Mo-99, the aqueous nitric acid solution of from about 0.01 M to about 2 M, containing a uranium concentration of from about 50 gU/L to about 350 gU/L, is evaporated under vacuum and/or through heating. The resultant solution is acidified with a suitable amount of nitric acid, and water if needed, to yield a solution concentration of nitric acid of from about 4M to about 8M, and a uranium concentration of from about 350 gU/L to about 650 gU/L. The temperature of this solution may be raised to ensure that all the uranium remains in solution. This solution is then evaporated under reduced pressure and/or cooled in order to promote conditions suitable for the formation of crystals of uranyl nitrate hydrates from the solution. An example of such a uranyl nitrate hydrate is UO2(NO3)2.6H2O. The crystals are then separated from the supernatant that remains and can be washed with nitric acid. Most of the uranium from any solid uranium target suitable for the production of Mo-99 that can be dissolved, and then converted into a solution containing aqueous nitric acid of concentration from about 0.01 M to about 2 M and uranium of concentration from about 50 gU/L to about 350 gU/L, can be recovered using this crystallization process. Examples of suitable solid uranium targets include, but are not limited to, uranium metal foils, U3Si2 plates, UAlx targets and UO2 targets. Through dissolution and subsequent chemical processing of the solid targets, a solution of irradiated uranium (50-350 gU/L) in nitric acid (0.01-2 M) can be prepared for recovery of Mo-99. After recovery of the majority of the Mo-99 using a sorbent, the remaining solution can be conditioned for crystallization of uranyl nitrate hydrates. Crystallization of uranyl nitrate hydrates removes most of the uranium from solution. The crystals can be filtered or otherwise removed from the supernatant and washed with nitric acid. Only a small fraction of the U-235 component of the uranium undergoes fission during irradiation. Removal of the Mo-99 along with other fission products with the sorbent separation process provides a uranium-containing solution having a greatly reduced amount of fission products. Additionally, many fission products will remain soluble during uranium nitrate hydrates crystallization; including Ba-140, Zr-95, Ru-103 and Ce-141, and these fission products will thus be separated from uranium nitrate hydrates. Therefore, according to the present process, recovery of such a purified uranium product, as uranium nitrate hydrates, affords uranium for reuse, storage, or disposal. In the case of disposal, purification of the uranium nitrate hydrates reduces the hazardous nature of any eventual uranium waste form. Nitric acid that is used in the process may also be recovered. Thus, nitric acid can also be recycled, further minimizing hazardous waste. An embodiment process will allow (1) recovery of Mo-99 using a sorbent and (2) recovery of purified uranium from the irradiated target for reuse, storage, or disposal. The concentrations of fission products and other impurities in the crystallized uranium nitrate hydrates may be too high for reuse, storage, or disposal. In this case further purification of the uranium nitrate hydrates crystalline material can be undertaken. Additional purification can be accomplished by a number methods including washing the crystals with nitric acid, heating the crystals to sweat out impurities prior to washing, and/or undertaking a second recrystallization process. In the latter case the uranyl nitrate hydrates solid would be dissolved in nitric acid, and the resulting solution would be conditioned to yield a 350-650 gU/L solution in a nitric acid concentration of between 4-8 M prior to crystallization through concentration by evaporation under reduced pressure and/or by cooling. 80% or greater of the Mo-99 produced from the U-235 fission in a solid uranium target (not corrected for radioactive decay) may be recovered after a sorbent-based separation, and 93% or greater of the uranium may be recovered in a purified form. After the solid target irradiation and dissolution, a crude Mo-99 product is separated from the uranium using a sorbent. Additional purification steps on the crude Mo-99 will result in a pure Mo-99 product for use in Tc-99m generators. In an embodiment, a solution of uranium in nitric acid may be concentrated through evaporation and acidified to a concentration of nitric acid of between 4 M and 8 M and uranium in an amount of, for example, 500 gU/L. Cooling to a temperature effective for crystallization, forming crystals of uranyl nitrate hydrates, an effective temperature being a temperature of from about 10° C. to about −30° C. (e.g., −10° C.) allows crystallization of 93% or greater of the uranium as uranyl nitrate hydrates, which is a largely insoluble salt at such cold temperatures. Evaporation under reduced pressure may be used as a means of both cooling the solution and lowering solution volume to increase the percentage of uranyl nitrate hydrates crystallized from solution. The crystals of uranyl nitrate hydrates are filtered from the supernatant that remains. An inorganic oxidant may be added to the solution of irradiated uranium (50-350 gU/L) in nitric acid (0.01-2 M) to ensure all of the Mo-99 is in the +VI oxidation state. +VI is the preferred oxidation state for the separation of Mo-99 from the uranium in nitric acid, in the sorbent separation step. Suitable inorganic oxidants include potassium permanganate, hydrogen peroxide, and sodium persulfate. In another embodiment, a uranium solution could be irradiated instead of a solid target to generate Mo-99. In this case the solution containing irradiated uranium can be conditioned to produce a solution of uranium (50-350 gU/L) in nitric acid (0.01-2 M) suitable for sorbent recovery of Mo-99. After recovery of the majority of the Mo-99 using a sorbent, the remaining solution can be conditioned for crystallization of uranyl nitrate hydrates. The purified uranium nitrate hydrates from the irradiated uranium solution is then available for reuse, storage, or disposal. FIG. 1 provides a flow diagram for an embodiment process. The boxes refer to a particular material and the numbers 1 through 5, which are in between boxes refer to process steps. Thus, the topmost box refers to an irradiated solid target of enriched uranium. After target irradiation, step 1 refers to the irradiated uranium target dissolution, and conditioning to form an aqueous nitric acid solution having a concentration of from about 0.01 M to about 2 M (e.g., 0.5 M). The concentration of uranium would be from about 50 gU/L to about 350 gU/L. Next, process step 2 involves removal of greater than 80% of the Mo-99 (not corrected for radioactive decay) from the solution using a solid sorbent-based separation process. >98% of the uranium remains in the nitric solution and is subjected to process step 3. Process step 3 involves conditioning the solution by increasing the concentration of uranium nitrate to a concentration of from about 350 gU/L to about 650 gU/L and increasing the concentration of nitric acid to a concentration of from about 4 M to about 8 M. These results may be achieved by evaporation using heat and/or evaporation under a reduced pressure and addition of nitric acid. This solution may be held at above ambient temperature (e.g., 40° C.) to be sure all of the uranium is dissolved. Process step 4 is performed on the now more concentrated solution, and results in crystallization to form crystals of uranyl nitrate hydrates, and a supernatant. The uranyl nitrate hydrates contain greater than 93% of the uranium. The supernatant contains less than 7% of the uranium which can be subsequently recovered, if required. Process step 5 results in uranium for reuse, storage, or disposal. The aforementioned embodiments relate to the irradiation of solutions of uranium and subsequent recovery of Mo-99 for generating Tc-99m, and thus relate to satisfying an objective of using LEU for generating Mo-99 and subsequent reuse, disposal, or storage of the LEU. Although the present invention has been described with reference to specific details, it is not intended that such details should be regarded as limitations upon the scope of the invention, except as and to the extent that they are included in the accompanying claims. |
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claims | 1. A translating section for allowing translational movement in an X direction and a Y direction while preventing any substantial movement of the translating section in a Z direction, the translating section comprising:(A) a frame having first and second mounting surfaces;(B) at least two actuators having first and second ends, wherein each of the actuators is extendable and retractable, and the first end of each respective actuator is attached to the corresponding first and second mounting surfaces of the frame respectively;(C) a center translation platform attached between the second ends of the actuators; and(D) a plurality of flexures, a flexure arranged between each of said second actuator ends and said center translation platform to allow translational movement of the center translation platform when the actuators extend and retract during scanning operation. 2. The translating section of claim 1, wherein each actuator further comprises a pair of piezoelectric elements equal in length. 3. The translating section of claim 1, wherein a flexure is arranged between each of said first actuator ends and said first and second mounting surfaces of the frame. |
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claims | 1. A beam forming system comprising one or more beam forming elements that are arranged to provide a non-planar doubly ruled radiation surface, the surface being defined by two families of rulings and wherein the length of the rulings within each family are configured to provide a radiation surface with substantially straight boundary edges, and wherein the beam forming system is arranged to form acoustic beams. 2. A beam forming system according to claim 1 wherein the radiation surface is a hyperbolic paraboloid. 3. A beam forming system according to claim 1 wherein the radiation surface is a hyperboloid of one sheet. 4. A beam forming system according to claim 1 wherein the beam forming elements comprise an array of straight elongate beam forming elements that are arranged relative to each other to collectively provide the non-planar doubly ruled radiation surface. 5. A beam forming system according to claim 4 wherein each straight elongate beam forming element comprises a plurality of beam forming elements that act together to from the straight elongate beam forming element. 6. A beam forming system according to claim 4 wherein the straight elongate beam forming elements are fixed relative to each other within a support structure so as to provide a predetermined beam-width of beams propagating from the radiation surface. 7. A beam forming system according to claim 4 wherein the straight elongate beam forming elements are movably supported relative to each other within a frame system, the frame system being operable to rotate the straight elongate beam forming elements relative to each other to alter angular parameters of the non-planar doubly ruled radiation surface so as to vary the beam-width of beams propagating from the radiation surface. 8. A beam forming system according to claim 1 in which a single beam forming element provides the non-planar doubly ruled radiation surface. 9. A beam forming system according to claim 8 wherein the single beam forming element comprises a continuous sheet of beam forming material that conforms to a non-planar doubly ruled surface. 10. A method of producing an acoustic beam comprising driving a beam forming system of claim 1 with a uniform phase. 11. A beam forming system according to claim 1 that is arranged as an acoustic transducer for projecting and/or receiving acoustic beams from or at the radiation surface. 12. A beam forming system according to claim 11 wherein the or each beam forming element is an acoustic transducer element comprising an active acoustic material that is operatively driven by front and back electrodes provided on opposing surfaces of the active acoustic material. 13. A beam forming system according to claim 12 further comprising a control system that is operatively connected to the electrodes and is arranged to drive the electrodes with electrical signals to cause the acoustic transducer to project and/or receive acoustic beams. 14. A beam forming system according to claim 13 wherein the control system is configured in an active mode to drive the acoustic transducer element(s) with electrical signals having uniform phase to produce an acoustic beam from each acoustic transducer element such that the overall acoustic beam projected from the radiation surface is the superposition of all of the acoustic beams from each acoustic transducer element. 15. A beam forming system according to claim 12 wherein the front and back electrodes are aligned with one family of rulings of the doubly ruled radiation surface. 16. A beam forming system according to claim 12 wherein the front electrodes of the acoustic transducer element(s) are aligned with one family of rulings of the doubly ruled radiation surface and the back electrodes are aligned with the other family of rulings of the doubly ruled radiation surface to provide a matrix electrode network. 17. A beam forming system according to claim 16 wherein a control system is operatively connected to all the electrodes in the matrix electrode network, and is operable to selectively drive different combinations of electrodes with electrical signals to thereby drive different combinations of selected acoustic transducer elements to produce any of the following: a wide-angle acoustic beam, acoustic beam-stripe, or an acoustic spot-beam. 18. A beam forming system according to claim 1 wherein each beam forming element comprises passive acoustic material such that the beam forming system can operate as an acoustic reflector or acoustic diffuser of acoustic beams incident on its radiation surface. 19. A beam forming system according to claim 18 wherein the passive acoustic material of the beam forming element(s) is acoustically reflective. 20. A beam forming system according to claim 1 wherein the system is an active beam forming device. 21. A beam forming system according to claim 1 wherein the system is an active acoustic device comprising beam forming element(s) comprising acoustically active material. 22. A beam forming system according to claim 1 wherein the system is an acoustic reflector comprising beam forming element(s) that are straight elongate active or passive acoustic elements of reflective material. 23. A beam forming system according to claim 1 wherein the system is an acoustic diffuser comprising beam forming element(s) that are straight elongate active or passive acoustic elements of reflective material. 24. A beam forming system comprising one or more beam forming elements that are arranged to provide a non-planar doubly ruled radiation surface, wherein the beam forming system is arranged to form acoustic beams. |
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043550029 | claims | 1. In a nuclear fuel assembly including first nuclear fuel elements containing fissionable material and a non-fissionable burnable poison, and second adjacent nuclear fuel elements containing the same fissionable material as contained in said first nuclear fuel elements but being free of said non-fissionable burnable poison, the improvement comprising the enrichment of said fissionable material in said first nuclear fuel elements at zero megawatt.day/ton burn-up being less than the enrichment of said same fissionable material in said second adjacent nuclear fuel elements at zero megawatt.day/ton burn-up, whereby the danger of damage to said first nuclear fuel elements is reduced. 2. A nuclear fuel assembly according to claim 1, wherein the enrichment of said fissionable material in said first nuclear fuel elements is less than about 72% of the enrichment of said fissionable material in said second adjacent nuclear fuel elements. 3. A nuclear fuel assembly according to claim 2, wherein said nuclear fuel elements are in the form of axially elongated rods, and wherein the enrichment of fissionable material in the axially middle portions of said first nuclear fuel elements is less than the enrichment of fissionable material in the same portion of said second adjacent nuclear fuel elements. 4. A nuclear fuel assembly according to claim 3, wherein the enrichment of fissionable material in said axially middle portions of said first nuclear fuel elements is less than the enrichment of the same fissionable material in the axially end portions of said first nuclear fuel elements. 5. A nuclear fuel assembly according to claim 3, wherein the enrichment of fissionable material in said axially middle portions of said first nuclear fuel elements is less than about 72% of the enrichment of fissionable material in said second adjacent nuclear fuel elements. 6. A nuclear fuel assembly according to claim 5, wherein the enrichment of fissionable material in said axially middle portions of said first nuclear fuel elements is less than the enrichment of the same fissionable material in the axially end portions of said first nuclear fuel elements. 7. A nuclear fuel assembly according to claim 3, wherein said fissionable material contained in said first and second nuclear fuel elements is present in pellets of UO.sub.2, and wherein said non-fissionable burnable poison contained in said first nuclear fuel elements is gadolinium oxide. 8. A nuclear fuel assembly according to claim 1, wherein said first nuclear fuel elements contain a single fissionable material. 9. A nuclear fuel assembly according to claim 1, wherein said nuclear fuel elements are in the form of axially elongated rods, and wherein the enrichment of said fissionable material in the axially middle portions of said first nuclear fuel elements is less than the enrichment of said fissionable material in identical portions of said second adjacent nuclear fuel elements. 10. A nuclear fuel assembly according to claim 9, wherein the enrichment of said fissionable material in the axially middle portions of said first nuclear fuel elements is less than about 72% of the enrichment of said second adjacent nuclear fuel elements. 11. A nuclear fuel assembly according to claim 1, wherein said second adjacent nuclear fuel elements are arranged adjacently surrounding each of said first nuclear fuel elements. 12. A nuclear fuel assembly according to claim 1, wherein said fissionable material contained in said first and second nuclear fuel elements is present in pellets of UO.sub.2, and wherein said non-fissionable burnable poison contained in said first nuclear fuel elements is gadolinium oxide. 13. A nuclear fuel assembly according to claim 12, wherein the enrichment of said fissionable material in said first nuclear fuel elements is less than about 72% of the enrichment of said fissionable material in said second adjacent nuclear fuel elements. 14. A nuclear fuel assembly according to claim 1, wherein said nuclear fuel elements are in the form of axially elongated rods. 15. A nuclear fuel assembly according to claim 1, wherein said non-fissionable burnable poison is a material having a high thermal neutron absorption cross-section. 16. A nuclear fuel assembly according to claim 15, wherein said non-fissionable burnable poison is a material containing at least one of boron, cadmium, erbium, europium, hafnium, samarium, and chemical compounds thereof. 17. In a nuclear fuel assembly including first nuclear fuel elements containing fissionable material and a non-fissionable burnable poison, and second adjacent nuclear fuel elements containing the same fissionable material as contained in said first nuclear elements but being free of said non-fissionable burnable poison, a method for preventing failure of said first nuclear fuel elements comprising the step of loading fissionable material in said first nuclear fuel elements so that the enrichment of the fissionable material in said first nuclear fuel elements at zero megawatt.day/ton burn-up is less than the enrichment of the same fissionable material in said second adjacent nuclear fuel elements at zero megawatt.day/ton burn-up, whereby the danger of damage to said first nuclear fuel elements is reduced. 18. A method according to claim 17, wherein the enrichment of fissionable material loaded in said first nuclear fuel elements is less than about 72% of the enrichment of fissionable material in said second adjacent nuclear fuel elements. |
summary | ||
description | The present application is a continuation of U.S. application Ser. No. 11/985,269, filed Nov. 13, 2007, now U.S. Pat. No. 7,629,586 now U.S. Publication No. 2008-0111082 dated May 15, 2008, which Claims priority to and the benefit of U.S. Provisional Application No. 60/858,773 filed Nov. 10, 2006. The entire content of each of the above-referenced applications are incorporated herein by reference. The invention relates generally to multi-modality medical imaging. More particularly, the invention relates to methods and systems for combining magnetic resonance imaging (MRI) with single photon nuclear imaging, such as single photon emission computed tomography (SPECT). Magnetic resonance imaging is an imaging technique used to visualize the inside of an object (or subject) under study (e.g., a human or animal body or a body part or an entire laboratory animal or specimen from the animal or a plastic test phantom). MRI relies on the relaxation properties of excited hydrogen nuclei in water and fat. When the object to be imaged is placed in a powerful, uniform magnetic field, the spins of the atomic nuclei with non-zero spin numbers (essentially, an unpaired proton or neutron) within the tissue all align in one of two opposite directions: parallel to the magnetic field or antiparallel. Magnetic field strengths for MRI studies of animals typically require 4.7 T, and magnets up to 17 T have been reported. For a comparison, the average magnetic field of the Earth is around 50 μT (or 0.5 G). Single photon emission computed tomography (SPECT) is a nuclear medicine tomographic imaging technique using gamma-rays. Conventionally, this imaging technique accumulates counts of gamma photons that are absorbed by a scintillator crystal. The crystal scintillates in response to interaction with gamma radiation to produce a flash of light. Photomultiplier tubes (PMTs) behind the scintillator crystal detect the flashes of light and a computer sums the fluorescent counts. The sum of fluorescent counts is a measure of the energy of an individual detected gamma-ray, and the location of the detected gamma-ray is computed from the distribution of the fluorescent counts among several PMTs. The computer in turn constructs an image of the relative spatial density of gamma-ray counts, accumulated as a series of detected gamma-rays whose measured energy is within a range that is selected by the operator, and displays the image on a computer monitor. This image then reflects the distribution and relative concentration of radioactive tracer elements present in the organs and tissues imaged. Although there may be benefits to combine SPECT and MRI, any theoretical benefits of trying to combine SPECT and MRI within a single system have been mostly dismissed because the functions of the PMTs in a typical SPECT system are severely compromised by the high magnetic fields needed for MRI and because magnetic field uniformity needed for MRI is distorted by the PMTs (i.e., the ferro-magnets in the PMTs). Recent advances in semiconductor technology have opened the possibility of replacing the PMTs and the scintillator crystal of a SPECT system with a semiconductor detector, such as a cadmium zinc telluride (CdZnTe or CZT) detector. The CZT detector may operate in the magnetic field inside an MR imaging apparatus. The CZT detector is referred to as a direct detector of radiation and operates by producing negative and positive charges (or electrons and holes) through interaction with gamma photons. However, combining a CZT detector for detecting gamma photons is still not a trivial task because the electrons and holes of the CZT detector need to travel non-negligible distances to generate their signals (e.g., travel distances of 2-5 mm and even larger). This presents possible Lorentz-force effects where signal generation may be distorted. In addition, it may be necessary to remove the electronics for signal amplification, address generation, logical operations, and other processing functions from the CZT module (in the high magnetic field) and to bring these electronics to a more distant location (in which a lower field can be found), thereby removing a cause of interference (e.g., either the offending electronics does not function in the high field or the offending electronics causes the MRI to have artifacts). However, the electronics located away from the magnetic field need to be connected via relatively long cables that result in an increased signal noise and distortion. In view of the foregoing and as discussed in Wagenaar et al. “Rational for the Combination of Nuclear Medicine with Magnetic Resonance for Pre-clinical Imaging,” Technology in Cancer Research and Treatment, ISSN 1533-0346, 2006, Vol. 5, No. 4, pp. 343-350, which is incorporated by reference herein in its entirety, it would be desirable to combine MRI with single photon nuclear imaging, such as SPECT, to provide a more complete coverage between high resolution, anatomical imaging, and genetically targeted molecular imaging that overcomes the detrimental effect of the magnetic fields produced by the MRI. The above information disclosed in this Background section is only for enhancement of understanding of the background of the invention and therefore it may contain information that does not form the prior art that is already known in this country to a person of ordinary skill in the art. An aspect of the present invention provides a dual-modality, fused image dataset from MRI and single-photon nuclear medicine imaging modalities in a single imaging session. The single imaging session allows a body (e.g., a human or animal body) or other object being scanned to remain motionless for sequential scanning while using the same body position on the same bed, thereby minimizing mis-registration artifacts from changes in body orientation between imaging studies. The single session also allows the simultaneous operation of the two modalities, providing exact co-registration in both position as well as in time. The ability to perform fused dual-modality imaging is helpful in both clinical imaging as well as pre-clinical research studies involving humans or laboratory animals for the development of drugs and therapies or the general study of biological processes. A combined magnetic resonance and single photon nuclear imaging system according to an embodiment of the present invention includes at least one semiconductor detector, at least one collimator, at least one magnet, and at least one transceiver. The least one semiconductor detector is for detecting gamma photons. The at least one collimator is for single photon nuclear imaging of an object under study with the at least one semiconductor detector. The at least one magnet is for producing a magnetic field suitable for magnetic resonance imaging. The at least one transceiver is for magnetic resonance imaging the object under study with the at least one magnet. Here, the at least one semiconductor detector is configured to single photon nuclear image the object under study under an influence of the magnetic field suitable for magnetic resonance imaging. In one embodiment of the system, the at least one semiconductor detector includes a material selected from the group consisting of silicon (Si), germanium (Ge), cadmium telluride (CdTe), mercuric iodide (HgI2), thallium bromide (TlBr), gallium arsenide (GaAs), cadmium zinc telluride (CdZnTe), and cadmium manganese telluride (CdMnTe). In one embodiment of the system, the at least one semiconductor detector is a cadmium zinc telluride (CZT) detector. In one embodiment of the system, the at least one semiconductor detector includes at least one semiconductor substrate and a plurality of electrodes. The at least one semiconductor substrate is for producing charge carriers through interaction with gamma photons, and the plurality of electrodes is for collecting the charge carriers to determine the gamma-ray energy and for localizing the gamma-ray interaction. In one embodiment of the system, the at least one semiconductor detector includes a semiconductor detector ring. The semiconductor detector ring may include a plurality of semiconductor linear sides. Each of the plurality of semiconductor linear sides may include a plurality of semiconductor modules. The semiconductor detector ring may include a plurality of semiconductor module rings. The plurality of semiconductor module rings may include a first ring having a plurality of first modules and a second ring having a plurality of second modules. The first modules of the first ring may be aligned with the second modules of the second ring along an axial direction, or the first modules of the first ring may have an angular offset with the second modules of the second ring along an axial direction. In one embodiment of the system, the at least one collimator is configured to be positioned between the object under study and the at least one semiconductor detector. In one embodiment of the system, the at least one semiconductor detector is a stationary detector. In one embodiment of the system, the at least one magnet includes a central opening, and the at least one semiconductor is configured to single photon nuclear image the body at either end of the at least one magnet and outside the central opening of the at least one magnet such that the object under study is single photon nuclear imaged and magnet resonance imaged in a sequential manner. In one embodiment of the system, the at least one magnet includes a central opening, and the at least one semiconductor is configured to single photon nuclear image the object under study within the central opening such that the object under study is capable of being single photo imaged and magnet resonance imaged in a substantially simultaneous manner. In one embodiment of the system, the imaging system further includes a gradient coil attached to the at least one magnet. The at least one magnet may include a central opening, the at least one transceiver includes a radio frequency (RF) coil, and the at least one semiconductor is configured to be between the RF coil and the gradient coil. In addition, the at least one semiconductor may be attached to the gradient coil, or the at least one collimator may be attached to the RF coil. In one embodiment of the system, the imaging system further includes a correction processor, the at least one semiconductor detector includes at least one semiconductor substrate for producing electrons upon an interaction with gamma photons, and the correction processor is adapted to compensate for a Lorentz-force effect on the electrons traveling within the at least one semiconductor substrate and under the influence of the magnetic field suitable for magnetic resonance imaging such that a drift of the electrons is compensated. In one embodiment of the system, the at least one semiconductor detector is adapted to detect at least one of the gamma photons emitted by the object under study and to generate a direct detection signal in response, a signal processor is adapted to receive the detection signal and includes a plurality of electronics adapted to amplify, address, and process the detection signal, and the signal processor is positioned away from the magnetic field suitable for magnetic resonance imaging to remove an interference effect of the magnetic field suitable for magnetic resonance imaging. In another embodiment of the present invention, a method of combining magnetic resonance and single photon nuclear imaging is provided. The method includes: injecting a radioactive isotope into an object under study; detecting gamma photons from the radioactive isotope within the object under study by at least one semiconductor detector; single photon nuclear imaging the object under study with at least one collimator positioned between the object under study and the at least one semiconductor detector; producing a magnetic field suitable for magnetic resonance imaging by at least one magnet; and magnetic resonance imaging the object under study with at least one transceiver positioned between the object under study and the at least one magnet. Here, the object under study is single photon nuclear imaged under an influence of the magnetic field suitable for magnetic resonance imaging. In one embodiment of the method, the step of detecting the gamma photons further includes: interacting the gamma photons with at least one semiconductor substrate of the at least one semiconductor detector; and collecting charge carriers produced by the interaction of the gamma photons with the at least one semiconductor substrate. In one embodiment of the method, the at least one semiconductor detector includes a first modular ring having a plurality of first modules and a second modular ring having a plurality of second modules. Here, the object under study may be single photon nuclear imaged by aligning the first modules of the first modular ring with the second modules of the second modular ring along an axial direction. Alternatively, the object under study may be single photon nuclear imaged by angular offsetting the first modules of the first modular ring with the second modules of the second modular ring along an axial direction. In one embodiment of the method, the object under study is single photon nuclear imaged by not moving the at least one semiconductor detector. In one embodiment of the method, the at least one magnet includes a central opening, and the object under study is single photon nuclear imaged by the at least one semiconductor at either end of the at least one magnet and outside the central opening of the at least one magnet such that the object under study is single photon nuclear imaged and magnet resonance imaged in a sequential manner. In one embodiment of the method, the at least one magnet includes a central opening, and the object under study is single photon nuclear imaged by the at least one semiconductor within the central opening such that the object under study is capable of being single photon imaged and magnet resonance imaged in a substantially simultaneous manner. In one embodiment, the method further includes the step of correcting for a Lorentz-force effect on electrons traveling within at least one semiconductor substrate of the at least one semiconductor detector and under the influence of the magnetic field suitable for magnetic resonance imaging. In one embodiment, the method further includes the steps of generating at least one direct detection signal in response to detecting the gamma photons by the at least one semiconductor detector; and receiving the detection signal by a signal processor having a plurality of electronics adapted to amplify, address, and process the detection signal; and removing an interference effect of the magnetic filed suitable for magnetic resonance imaging on the single photon nuclear imaging by positioning the signal processor away from the magnetic field suitable for magnetic resonance imaging. In the following detailed description, only certain exemplary embodiments of the present invention are shown and described, by way of illustration. As those skilled in the art would recognize, the described exemplary embodiments may be modified in various ways, all without departing from the spirit or scope of the present invention. Accordingly, the drawings and description are to be regarded as illustrative in nature, and not restrictive. An embodiment of the present invention is designed to enhance the MRI imaging by incorporating an additional modality in the same gantry as operated by an MRI machine. The added modality is SPECT, limited-angle SPECT, or planar imaging based on the single photon emission principle. In one embodiment of the present invention, the single photon emission imagers are based on a semiconductor direct conversion detector, such as a cadmium zinc telluride (CZT) detector. The embodiment of the present invention reduces the possibility of operational mistakes. The embodiment avoids the changing of the position of the human or animal being imaged, and ensures the accuracy of the co-registration between the data acquired from the two modalities. It allows for the simultaneous acquisition of dynamic and/or static data sets and the single-injection of combined contrast agents for the two modalities. Moreover the single photon imager data are not detrimentally affected by the magnetic fields produced by the MRI scanner (or imager) and vice versa. In more detail, conventional nuclear medicine imaging relies on the use of PMTs to detect light flashes from the absorption of gamma-rays in scintillator crystals. As discussed above, the PMTs, however, do not work in magnetic fields. In one embodiment of the present invention, by replacing the scintillator and PMT combination with a solid-state semiconductor detector, such as a CdZnTe or CZT detector, the embodiment of the present invention realizes a gamma camera that can operate in the magnetic field inside an MR imaging apparatus. Multimodality imaging offers many opportunities for the combination of spatially and temporally-registered data. One embodiment of the present invention combines anatomical context and functional information, such as the anatomical delineation of the boundaries of a tumor (using, e.g., MRI) with the functional definition of aggressive cancer cells at the perimeter and necrotic cells at the core of the tumor (using, e.g., SPECT). This is but one of many possible combinations of imaging data, and the present invention is not thereby limited. In one embodiment of the present invention, the combination of MRI data with single-photon nuclear imaging data with spatial and temporal registration is realized through the use of the semiconductor nature of the CZT in order to overcome the magnetic field limitations of conventional PMTs (i.e., since this combination has been technologically precluded by the requirement of PMTs). Multi-modality imaging should not require a movement over a considerable distance of the object under study (e.g., the human or animal body being imaged), such that a single photon nuclear imaging device (e.g., a PMT detector) is at a considerable distance away from the presence of the magnetic field. That is, moving the animal or patient greatly increases the problem of co-registration of images from the two modalities. Also co-registered images may lose some of their precision if organs or body parts are located at different positions (i.e., they have shifted) during the imaging sessions. As such, one embodiment of the present invention includes a semiconductor CZT detector that can simultaneously or sequentially (in close proximity) provide single photon nuclear imaging (e.g., SPECT) and MRI imaging because the semiconductor CZT imaging detector can operate in a magnetic field, whereas the PMT-based imaging devices cannot create useful images in a magnetic field. That is, simultaneous imaging is possible because the single photon nuclear imaging (or SPECT) system of an embodiment of the present invention is located inside the field of the MRI system. Referring to FIG. 19, a semiconductor detector according to an embodiment of the present invention includes a semiconductor substrate (or crystal) 300 for producing charge carriers (electrons or holes) through interaction with gamma photons and electrodes (e.g., anodes or cathodes) 310 for collecting the charge carriers. In FIG. 19, the substrate 300 is made mainly of CZT. However, the present invention is not thereby limited, and a substrate of a semiconductor radiation detector can be made mainly of another compound semiconductor such as silicon (Si), germanium (Ge), cadmium telluride (CdTe), mercuric iodide (HgI2), thallium bromide (TlBr), gallium arsenide (GaAs), and cadmium manganese telluride (CdMnTe). The principle of operation of a semiconductor detector is the following: if a photon interacts within the detector, all or part of its energy is converted into the liberation of free electrons and holes, the number of electron-hole pairs being proportional to the photon energy converted in the interaction. An externally applied electric field separates the pairs before they recombine; electrons drift toward the anodes, which define the detector's pixels, holes to the cathode; the charge is collected by the electrodes (charge collection) 310. The collected charge produces a current pulse on the electrode 310, whose integral equals the total charge generated by the incident particle, i.e., is a measure of the deposited energy. The readout goes through a charge-sensitive preamplifier, followed by a shaping amplifier. One embodiment of the present invention includes pixellated semiconductor imaging modules made of CZT. However, the semiconductor imaging module does not necessarily have to be CZT, and it can be another compound semiconductor such as silicon (Si), germanium (Ge), cadmium telluride (CdTe), mercuric iodide (HgI2), thallium bromide (TlBr), gallium arsenide (GaAs), and cadmium manganese telluride (CdMnTe). In one embodiment, these modules are square and planar and can be tiled to form a line (i.e., a ladder) or a rectangular mosaic of modules. The semiconductor does not generally interrupt the operation of the MRI components, and the strong magnetic field does not generally disturb the functionality of the semiconductor detector. Having both modalities capable of simultaneous or adjacent and sequential imaging can thus be realized. In order to perform tomographic imaging, the semiconductor modules have to sufficiently sample in the angular direction. A ring of the modules is the most straightforward way to provide complete angular sampling for tomography, and FIGS. 1 through 10 depict various ring configurations pursuant to embodiments of the present invention. In more detail, FIG. 1 depicts a configuration of a combined magnetic resonance and single photon nuclear imaging system according to an embodiment of the present invention. As shown in FIG. 1, the imaging system includes a semiconductor detector 11 for detecting gamma photons and a collimator (described in more detail below) for single photon nuclear imaging of an object (or subject) under study (e.g., a human or animal body) with the semiconductor detector 11. In addition, the imaging system includes a magnet 14 for producing a magnetic field suitable for magnetic resonance imaging and a transceiver for magnetic resonance imaging of the object under study with the magnet 14. Here, the semiconductor detector 11 is configured to single photon nuclear image the object under study being imaged under the magnetic field suitable for magnetic resonance imaging. In FIG. 1, an “adjacent” configuration in which the semiconductor detector 11 is a CZT ring or a ring of semiconductor detectors (e.g., a SPECT ring) attached to the outside surface of the magnet 14 is shown. The patient (or animal or other object) is sequentially imaged, in the CZT ring of the semiconductor detector 11 first and then magnetic resonance imaged by the magnet 14, or vice-versa, with the patient (or animal or other object) on the same bed in different axial positions such that the animal is sequentially in the fields-of-view of the two respective instruments. FIG. 2 depicts another configuration of a combined magnetic resonance and single photon nuclear imaging system according to an embodiment of the present invention. In FIG. 2, the imaging system includes a semiconductor detector 21 for detecting gamma photons and a collimator for single photon nuclear imaging an object under study with the semiconductor detector 21. In addition, the imaging system includes a magnet 24 for producing a magnetic field suitable for magnetic resonance imaging and a transceiver 22 for magnetic resonance imaging the object under study with the magnets 24. Here, the semiconductor detector 21 is configured to single photon nuclear image the object under study under the magnetic field suitable for magnetic resonance imaging. In more detail, the semiconductor detector 21 is illustrated as a CZT ring attached to a gradient coil 23. The gradient coil 23 is attached to the magnet 24 and is seldom removed from the magnet 24. As such, the CZT ring of the semiconductor detector 21 is also seldom removed from the magnet 24 (i.e., the CZT ring is semi-permanent like the gradient coil 23). Here, the operation of the CZT and the MRI can be simultaneous or sequential, with simultaneous acquisition having certain advantages, as discussed in Wagenaar et al. “Rational for the Combination of Nuclear Medicine with Magnetic Resonance for Pre-clinical Imaging,” Technology in Cancer Research and Treatment, ISSN 1533-0346, 2006, Vol. 5, No. 4, pp. 343-350, which is incorporated by reference herein in its entirety. FIG. 3 depicts yet another configuration of a combined magnetic resonance and single photon nuclear imaging system according to an embodiment of the present invention. In FIG. 3, the imaging system includes a semiconductor detector 31 for detecting gamma photons and a collimator for single photon nuclear imaging of an object under study with the semiconductor detector 31. In addition, the imaging system includes a magnet 34 for producing a magnetic field suitable for magnetic resonance imaging and a transceiver 32 for magnetic resonance imaging of the object under study with the magnet 34. Here, the semiconductor detector 31 is configured to single photon nuclear image of the object under study under the magnetic field suitable for magnetic resonance imaging. In more detail, the transceiver 32 is illustrated as an RF coil, the semiconductor detector 31 is illustrated as a CZT ring coupled to the RF coil via the collimator (described in more detail below), and a gradient coil 33 is shown to be attached to the magnet 34. Here, the collimator is attached to the RF coil and is used only in special cases when the particular RF coil is used. That is, the collimator is attached or is very close to the RF coil because the collimator is between the CZT ring (semiconductor detector) and the RF coil. As such, the CZT ring and/or the collimator is(are) removable like the RF coil (i.e., not semi-permanent). Here, like the embodiment of FIG. 2, the operation of the CZT and the MRI can be simultaneous or sequential, with simultaneous acquisition having certain advantages, as discussed in Wagenaar et al. “Rational for the Combination of Nuclear Medicine with Magnetic Resonance for Pre-clinical Imaging,” Technology in Cancer Research and Treatment, ISSN 1533-0346, 2006, Vol. 5, No. 4, pp. 343-350, which is incorporated by reference herein in its entirety. Referring to FIGS. 4 and 5, a semiconductor detector in accordance with an embodiment of the present invention is composed of a semiconductor detector ring, such as a CZT ring. The semiconductor ring, such as the CZT ring, includes a plurality of semiconductor linear sides (e.g., linear sides of CZT) M. In FIG. 4, semiconductor detector rings respectively having three (3), four (4), five (5), six (6), and eight (8) sides M are shown. In FIG. 5, semiconductor detector rings respectively having 12 and 24 sides M are shown. The present invention, however, is not limited to the side numbers shown in FIGS. 4 and 5. For example, the number of sides M can be any positive number ranging from 1 to 128. Referring now to FIG. 6, each of the semiconductor linear sides (or linear sides of CZT) M can have one or more individual semiconductor modules P. In FIG. 6, semiconductor linear sides M respectively having 1, 2, and 3 semiconductor modules P are shown. Also, referring to FIG. 7, a semiconductor detector in accordance with an embodiment of the present invention can also be composed of a number of semiconductor module rings, such as a number of CZT rings, N. In FIG. 7, a side of semiconductor detectors respectively having 1, 2, 3, and 4 rings N are shown. To put it another way, the semiconductor (or CZT) modules P are assembled on linear “ladders,” linear in the sense that they are a line of semiconductor (or CZT) modules P as shown in FIG. 7. In FIG. 7, the numbers of semiconductor modules P in the ladders are respectively given as 1, 2, 3, and 4. In one embodiment, ladders such as shown in FIG. 7 with N=4 are typically arranged into a cylindrical “barrel” in high energy physics experimental setups, with the number of ladders “M” being the number of sides of the barrel detector surrounding the high energy physics experimental interaction volume. Referring to FIG. 8, each of the semiconductor linear sides M can also be a certain number of modules P wide. In FIG. 8, a side M is shown to be 4 modules P wide and having 3 rings N. Referring to FIG. 9, the “coverage” of a semiconductor module ring (e.g., a semiconductor module ring N) can be complete, as shown in the sides M of the polygons of FIGS. 4 and 5, or partial as shown in FIG. 9. That is, a polygon having a plurality of sides M can be populated by CZTs in any number of sides from 1 to the total number of the sides M as shown in FIG. 9. However, it is noted that partial coverage of the polygon can result “limited angle tomography” in the field of nuclear medicine imaging. Limited angle tomography does not cover the complete range of angular sampling. Modules P (or sides M) of a first semiconductor module ring (or one ring of modules) can be aligned with modules P (or sides M) of a second semiconductor module ring (or a second ring of modules) along an axial direction as shown in FIGS. 7 and 8. Alternatively, as shown in FIG. 10, modules P (or sides M) of a first semiconductor module ring (or one ring of modules) 50 can have an angular offset with modules P (or sides M) of a second semiconductor module ring (or a second ring of modules) 60 along an axial direction. In FIG. 10, the offset of the first semiconductor module ring 50 and the second semiconductor module ring 60 allows these rings to view an object being imaged with twice the angular sampling with one axial motion. That is, as shown in FIG. 10, the object under study is single photon nuclear imaged by angular offsetting the modules P of the first modular ring 50 with the modules P of the second modular ring 60 along an axial direction so that these rings can view the object under study with twice the angular sampling. An image formation apparatus, such as a collimator, is positioned between the object being imaged and a CZT ring. As an example, in nuclear medicine, this image formation apparatus can be either a parallel hole collimator or a pinhole collimator. FIG. 11 depicts a single photon nuclear imaging system pursuant to an embodiment of the present invention. As shown in FIG. 11, the imaging system includes a four head detector (or detector heads) 100 with four pinhole collimators between the detector 100 and the object (e.g., a small animal) being imaged. Here, each of the four pinhole collimators includes a pinhole 120 and a pyramid-shaped cone 110 made of lead that suspend the pinhole 120 near the object being imaged. That is, the detector heads 100 are shielded from radiation by the pyramid-shaped cones 110 of lead that suspend the pinholes 120 near the object being imaged to single photon nuclear image the object being imaged with the detector heads 100. The present invention, however, is not limited to an imaging formation apparatus having one or more pinholes (or a pinhole collimator). That is, an image formation device can be composed of a single pinhole per CZT module, a single pinhole per CZT side (illuminating all rings), multiple pinholes per CZT side (illuminating all rings), a single slit per CZT side per ring, multiple slits illuminating a CZT side (all rings), slits that are either parallel to the axis of the MR field or tangentially oriented, a coded aperture array of holes that is formed in the cylinder or suspended between the object being imaged and the CZT side (all rings), a cylindrical coded aperture array, also known as a “ring coded aperture” that is positioned between the object being imaged and the CZT side (all rings), an array of stationary holes that is positioned between the object being imaged and the CZT ring (i.e., a collimator, either parallel hole or converging), and/or an array of stationary attenuating pins that is positioned between the object being imaged and the CZT ring (i.e., an inverse collimator with either parallel pins or converging pins). Also, in one embodiment of the present invention, the object under study is single photon nuclear imaged by not moving the semiconductor detector. That is, the image formation device (e.g., the collimator) can be completely stationary (i.e., non-moving) and its orientation relative to the CZT ring fixed. For example, the “coverage” of a semiconductor module ring (e.g., a semiconductor module ring N or detector heads) can completely surround an object being imaged so that the image formation device does not have to rotate (i.e., can be stationary) to view the complete (or entire) object as, e.g., shown in FIGS. 4 and 5. Further, a stationary detector can also overcome any eddy currents (or noises or distortions) that may be produced by a detector that rotates in a magnetic field. However, the present invention is not thereby limited and, alternatively, the image formation device (e.g., the collimator) can be composed of a cylinder that can rotate. Moreover, in one embodiment of the present invention, the image formation material (e.g., lead) of the image formation device should be highly absorbent of x-rays and gamma-rays but still minimally disturbing of the magnetic field. The rotating cylinder should be centered on the axis of the magnetic field, and the spatial distribution of mass should substantially remain unchanged during rotation, such that the magnetic field is minimally perturbed by the rotational movement. FIG. 12 depicts a configuration of a combined magnetic resonance and single photon nuclear imaging system according to an embodiment of the present invention. In more detail, FIG. 12 depicts an external dual-head pinhole MR-SPECT configuration or “adjacent” configuration in which a dual-headed pinhole SPECT system 250 is attached external to (or to the outside surface of) a magnet 260 of an MRI system. The patient (or animal) is sequentially imaged, by the pinhole SPECT system 250 first and then magnetic resonance imaged by the magnet 260, or vice-versa, with the patient (or animal) on the same bed in the same axial position. Here, the one or more detectors (or detector heads) of dual-headed pinhole SPECT system 250 can rotate. FIG. 13 shows gamma-ray images 400, 410, and 420 of a line phantom taken with a CZT camera. The image 400 was taken in the Earth field. The image 410 was taken at the location where the field was 50 Gauss. The image 420 was taken at the location where the field was 100 Gauss. Here, the line phantom has line spaces of 3.5 mm in the left section, top section 3 mm, right section 2.5 mm, and bottom section 2 mm. The pixel size of the gamma camera was 2.5 mm. Field direction was pointing from top to bottom. By contrast, FIG. 14 shows images 500, 510, 520, and 530 of five capillary tubes filled with 99 mTc solutions in the magnetic field, taken with a PSPMT gamma camera. Data in the 99 mTc peak were used to construct the images. The tube inner diameter was 0.5 mm, and image sizes are 12.5×12.5 cm2. The image 500 was taken in the Earth field, and the images 510, 520, and 530 were taken with a magnet 6 cm, 4.5 cm, and 3 cm from the lower edge, respectively. In images 510, 520, and 530 contour lines show field levels in Gauss. As such, the images of FIGS. 13 and 14 show the CZT camera to be insensitive to magnetic fields up to 100 G (FIG. 13), whereas the PMT-based system failed at about 15 G or 1.5 mT (FIG. 14). FIG. 15 demonstrates the complementary relationship between MRI and nuclear imaging, such as SPECT. In particular, when the two modalities are combined according to an embodiment of the present invention, the embodiment can realize complete coverage of the space between high resolution, anatomical imaging, and genetically targeted molecular imaging. FIG. 16 illustrates a method of combining magnetic resonance and single photon nuclear imaging in accordance with an embodiment of the present invention. In step 1000 of this embodiment, a radioactive isotope is injected into an object under study. Gamma photons from the radioactive isotope within the object under study are detected by at least one semiconductor detector in step 1010. In step 1020, the object under study is single photon nuclear imaged with at least one collimator positioned between the object under study and the at least one semiconductor detector. In one embodiment of the present invention, the object under study is single photon nuclear imaged by not moving the at least one semiconductor detector. Also, in one embodiment, the at least one semiconductor detector includes a first modular ring and a second modular ring, and the object under study is single photon nuclear imaged by angular offsetting the modules of the first modular ring with the modules of the second modular ring along an axial direction so that these rings can view the object under study with twice the angular sampling with one axial motion. Moreover, simultaneously, concurrently, or sequentially with the above steps 1000, 1010, and 1020, a magnetic field suitable for magnetic resonance imaging is produced by at least one magnet in step 1030, and the object under study is magnetic resonance imaged with at least one transceiver positioned between the object under study and the at least one magnet in step 1040. Here, in the method of FIG. 16, the object under study is single photon nuclear imaged under an influence of the magnetic field suitable for magnetic resonance imaging. In one embodiment, and referring to FIG. 17, a correction processor 600 is provided to the at least one semiconductor detector (or detectors). Here, the semiconductor detector has at least one semiconductor substrate (or substrates) for producing electrons upon an interaction with gamma photons, and the correction processor 600 is adapted to compensate for a Lorentz-force effect on the electrons traveling within the at least one semiconductor substrate and under the influence of the magnetic field suitable for magnetic resonance imaging such that a drift of the electrons is compensated. Referring to FIG. 18, the at least one semiconductor detector (or detectors) according to one embodiment of the present invention is adapted to detect one or more gamma photons emitted by the object under study and to generate a direct detection signal in response. In this embodiment, a signal processor 700 is coupled to the at least one semiconductor detector and adapted to receive the detection signal. The signal processor 700 includes a plurality of electronics adapted to amplify, address, and process the detection signal, and the signal processor 700 is shown in FIG. 18 to be positioned away from the magnetic field suitable for magnetic resonance imaging to remove an interference effect of the magnetic field suitable for magnetic resonance imaging. It should be appreciated from the above that the various structures and functions described herein may be incorporated into a variety of apparatuses (e.g., an imaging device, a monitoring device, etc.) and implemented in a variety of ways. Different embodiments of the imaging and/or monitoring devices may include a variety of hardware and software processing components. In some embodiments, hardware components such as processors, controllers, state machines and/or logic may be used to implement the described components or circuits. In some embodiments, code such as software or firmware executing on one or more processing devices may be used to implement one or more of the described operations or components. In view of the foregoing, some embodiments of the invention described herein generally relate to an apparatus and method for providing a dual-modality, fused image dataset from MRI and single-photon nuclear medicine imaging modalities in a single imaging session. The single imaging session allows an object (e.g., a human or animal body or a body part or an entire laboratory animal or specimen from the animal or a plastic test phantom) being scanned to remain motionless for sequential scanning while using the same body position on the same bed, thereby minimizing mis-registration artifacts from changes in body orientation between imaging studies. The single session also allows the simultaneous operation of the two modalities, providing exact co-registration in position as well as in time. The ability to perform fused dual-modality imaging is helpful in both clinical imaging as well as pre-clinical research studies involving laboratory humans or animals for the development of drugs and therapies or the general study of biological processes. While the invention has been described in connection with certain exemplary embodiments, it is to be understood by those skilled in the art that the invention is not limited to the disclosed embodiments, but, on the contrary, is intended to cover various modifications included within the spirit and scope of the appended claims and equivalents thereof. |
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041742575 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT A reactor vessel body 2 and a reactor vessel head 4 are joined by bolted connection at flange 6. The reactor vessel body has an inlet opening 8 and an outlet opening 10 for flow of coolant water therethrough. A core 12 is comprised of a plurality of fuel assemblies 14, each of which is comprised of a plurality of elongated fuel rods. The core is supported on the core support assembly 16 which is in turn supported by the core support barrel 18. This core support barrel is supported by flange 20 from the reactor vessel body 2 at a location adjacent the flange 6. Immediately above the core 12 is a fuel assembly alignment and seal plate 22 which serves to engage the upper ends of the fuel assemblies and to maintain alignment thereof. A boundary plate structure 24 is located above the alignment plate, thereby defining the outlet plenum 26. After the coolant enters through inlet opening 8 the flow passes downwardly through the annular space 28 between the reactor vessel and the core support barrel. This flow passes downwardly through the flow skirt 30 into an inlet plenum 32 located below the core 12. The flow passes upwardly through the core and through openings 52 in the alignment plate 22 into the outlet plenum 26. From here the flow passes out through outlet opening 10 to a steam generator (not shown). Each of the fuel assemblies 14 contain within their structure four control rod guide tubes 40 which pass through the entire length of the fuel assembly. Finger shaped control rods 48 are vertically movable within the guide tubes 40 of the fuel assemblies. Each of these rods individually extends to an elevation above the foundary plate 24 at which location they may be joined in subgroupings to the control rod extension 50. In addition to the flow holes 52, the alignment and seal plate 22 also has openings 54 through which the control rods pass. Control rod shroud tubes 56 pass through the outlet plenum 26 and may be welded to the alignment and seal plate 22 and the boundary plate structure 24. These shroud tubes surround and protect the control rods from the effects of cross flow through the plenum 26, and also are open to a chamber 72 above the boundary plate 24. The boundary plate 24 is supported from barrel 60 which is supported by flanges 62 resting on flanges 20 of the core support barrel. The upper guide structure support plate 64 is open to permit flow therethrough. The upper end of the control rod shroud tube 56 is open to the chamber 72 from which water passes by plate structure 24 to the outlet 10. Guide tubes 40 in the fuel assemblies are open at their upper end and therefore exposed to a low pressure existing near outlet 10. Referring to FIG. 2, the pressure at the top chamber of guide tubes 40 approximates the outlet pressure from the reactor. The control rod guide tubes 40 may be terminated just below the alignment and seal plate 22 as illustrated, or they may be continued upwardly into the top chamber 72. In either case, the tubes are open to a fluid pressure which approximate the outlet pressure from the reactor. The lower end of the guide tubes 40 extend through the seal plate 90 and are exposed to the pressurizable chamber 91. Slots 92 are provided in the core support structure 93 and the pressure plate 90 has downwardly extending lips 94 extending into slots in sealing relationship. Since the main flow of coolant is upwardly across the fuel assemblies 14, the pressure at the lower end of the fuel assemblies is higher than the pressure at the upper end. This high inlet pressure operates on the upper surface of the pressure plate 90. Depending on the amount of flow past the seal, openings 95 may be provided through the seal plate structure, to permit additional flow to pass into chamber 91 and up through control rod guide tubes 40 in a total amount sufficient to cool the control rod fingers. A substantial amount of the flow restriction is provided in these openigns 95 so that the pressure in the pressurizable plenum 91 approximates that at the outlet of the control rod guide tubes. It follows that a low pressure exists below the seal plate 90 and a relatively high pressure exists above the seal plate. This pressure differential operates to hold-down the fuel assembly 14. Supplementary springs 96 may be provided between the upper portion of the fuel assembly and the alignment and seal plate 22. A flow opening 99 may be provided as a flow path for the main coolant flow to the core. The low pressure chamber 91 is sealed around the periphery of the opening by downwardly extending lips 97 which engage the edges of slot 98 in the support and seal plate 93. The uplift forces on the fuel assemblies are a function of the flow through the core. In this invention the hold-down force is also a function of the flow through the core, and therefore, the forces are self-compensating. This provides more tolerance in the event that flow or pressure drop varies form that predicted. |
claims | 1. A detection apparatus that can determine whether spent nuclear fuel rods are missing or have been replaced with dummy spent nuclear fuel rods or have been replaced with fresh spent nuclear fuel rods, wherein the spent nuclear fuel rods are part of a spent nuclear fuel array located in a spent nuclear fuel storage pool, the spent nuclear fuel array having the spent nuclear fuel rods arranged in a quadrant symmetric pattern with guide tubes adjacent and between the spent nuclear fuel rods, comprising:a detection apparatus insertion fixture positioned in said spent nuclear fuel storage pool and adapted to contact the spent nuclear fuel array,said detection apparatus insertion fixture including a slider assembly with insertion tubes adapted to be inserted into the guide tubes in the spent nuclear fuel array,said insertion tubes moveable in the guide tubes of the spent nuclear fuel array by said slider assembly,a first detector and a second detector contained within the insertion tubes,wherein said first detector and said second detector are neutron detectors or gamma detectors and wherein said first detector and said second detector detect radiation responses,wherein the insertion tubes are between the spent nuclear fuel rods that are arranged in a quadrant symmetric pattern,wherein the insertion tubes containing said first detector and said second detector are inserted into the guide tubes adjacent the spent nuclear fuel rods and move through the guide tubes in the spent nuclear fuel array wherein said first detector and said second detector are in predetermined positions in the quadrant symmetric pattern;a measuring and analyzing device, said measuring and analyzing device connected to said first detector and said second detector wherein said first detector and said second detector are located in predetermined positions in the quadrant symmetric pattern for measuring said radiation responses of said first detector and said second detector at the predetermined positions in the quadrant symmetric pattern simultaneously at said predetermined positions in the quadrant symmetric pattern and processing said radiation responses and producing a signature; andwherein said measuring and analyzing device uses said signature for determining whether some spent nuclear fuel rods within the spent nuclear fuel array in the spent nuclear fuel storage pool are missing or have been replaced with dummy spent nuclear fuel rods or have been replaced with fresh spent nuclear fuel rods using said signature by determining whether said signature has been perturbed from the quadrant symmetric pattern. 2. The detection apparatus of claim 1 wherein said first detector and said second detector are neutron detectors. 3. The detection apparatus of claim 1 wherein said first detector and said second detector are gamma detectors. 4. The detection apparatus of claim 1 wherein said first detector is a neutron detector and said second detector is a gamma detector. 5. The detection apparatus of claim 1 wherein said measuring and analysis device includes a multi-channel analyzer and a computer. 6. The detection apparatus of claim 1 wherein the spent nuclear fuel array has alignment holes and wherein said detection apparatus insertion fixture includes guide pins adapted to engage the alignment holes in the spent nuclear fuel array. 7. The detection apparatus of claim 1 wherein the spent nuclear fuel array has alignment holes and wherein said detection apparatus insertion fixture includes guide pins adapted to engage the alignment holes in the spent nuclear fuel array and guide bushings for the insertion tubes, wherein the insertion tubes are adapted to move in said guide bushings. 8. A detection system that can determine whether spent nuclear fuel rods are missing or whether some spent nuclear fuel rods have been replaced with dummy spent nuclear fuel rods or fresh spent nuclear fuel rods, the detection system comprising:a spent nuclear fuel storage pool,a spent nuclear fuel array located in said spent nuclear fuel storage pool,said spent nuclear fuel array including the spent nuclear fuel rods arranged in a quadrant symmetric pattern and guide tubes adjacent and between the spent nuclear fuel rods,a detection apparatus insertion fixture adapted to be inserted into said spent nuclear fuel storage pool,said detection apparatus insertion fixture including a slider assembly with insertion tubes,said insertion tubes adapted to be moved in said guide tubes of said spent nuclear fuel array by said slider assembly in said spent nuclear fuel storage pool,wherein said insertion tubes contain a first detector and a second detector,wherein said first detector and said second detector are neutron detectors or gamma detectors and wherein said first detector and said second detector detect radiation responses,said insertion tubes are adapted to be inserted into said guide tubes adjacent the spent nuclear fuel rods in said spent nuclear fuel array through said guide tubes in said spent nuclear fuel array in predetermined positions in said quadrant symmetric pattern; anda measuring and analysis unit connected to said first detector and said second detector wherein said first detector and said second detector are located in predetermined positions in said quadrant symmetric pattern wherein radiation responses of said first detector and said second detector are simultaneously measured at a location or multiple locations within said guide tubes to produce a signature, anda processor for processing said radiation responses and said signature by determining whether said signature has been perturbed in said quadrant symmetric pattern and determining whether said spent nuclear fuel rods within said spent nuclear fuel array in said spent nuclear fuel storage pool are missing or have been replaced with dummy spent nuclear fuel rods or fresh spent nuclear fuel rods. 9. The detection apparatus of claim 8 wherein said first detector and said second detector are neutron detectors. 10. The detection apparatus of claim 8 wherein said first detector and said second detector are gamma detectors. 11. The detection apparatus of claim 8 wherein said first detector is a neutron detector and said second detector is a gamma detector. 12. The detection apparatus of claim 8 wherein said measuring and analysis unit includes a multi-channel analyzer and a computer. 13. A method of determining possible diversion of spent nuclear fuel rods located in a spent nuclear fuel storage pool in a spent nuclear fuel array in the spent nuclear fuel storage pool, wherein the spent nuclear fuel array includes spent nuclear fuel rods arranged in a fuel rods quadrant symmetric pattern and guide tubes adjacent and between the spent nuclear fuel rods, comprising the following steps:the step of providing a detector instrument insertion fixture having a slider assembly with insertion tubes arranged in a detector quadrant symmetric pattern corresponding to the fuel rods quadrant symmetric pattern, said insertion tubes containing a first detector and a second detector, wherein said first detector and said second detector are gamma ray detectors and wherein said first detector and said second detector are in predetermined positions in said detector quadrant symmetric pattern,the step of moving said detector instrument insertion fixture into the spent nuclear fuel storage pool adjacent the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of moving said slider assembly with said insertion tubes into contact with the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of inserting said insertion tubes containing said gamma ray detectors of said first detector and said second detector into said guide tubes in the spent nuclear fuel array using said detector instrument insertion fixture,the step of measuring gamma ray radiation responses of said gamma ray detectors in said guide tubes,the step of processing said gamma ray radiation responses in said guide tubes by normalizing them to the maximum value among them and producing a signature based on these normalized values, andthe step of producing an output that consists of said signature and determining whether said signature has been perturbed in said fuel rods quadrant symmetric pattern that can indicate possible diversion of said spent nuclear fuel rods from the spent nuclear fuel array in the spent nuclear fuel storage pool. 14. A method of determining possible diversion of spent nuclear fuel rods located in a spent nuclear fuel storage pool in a spent nuclear fuel array in the spent nuclear fuel storage pool, wherein the spent nuclear fuel array includes spent nuclear fuel rods arranged in a fuel rods quadrant symmetric pattern and guide tubes adjacent and between the spent nuclear fuel rods, comprising the following steps:the step of providing a detector instrument insertion fixture having a slider assembly with insertion tubes arranged in a detector quadrant symmetric pattern corresponding to said fuel rods quadrant symmetric pattern, said insertion tubes containing a first detector and a second detector, wherein said first detector and said second detector are neutron detectors,the step of moving said detector instrument insertion fixture into the spent nuclear fuel storage pool adjacent the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of moving said slider assembly with insertion tubes into contact with the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of inserting said insertion tubes containing said neutron detectors of said first detector and said second detector into the guide tubes in the spent nuclear fuel array using said detector instrument insertion fixture,the step of measuring neutron radiation responses of said neutron detectors in said guide tubes,the step of processing said neutron radiation responses in said guide tubes by normalizing them to the maximum value among them and producing a signature based on these normalized values, andthe step of producing an output that consists of said signature and determining whether said signature has been perturbed in said detector quadrant symmetric pattern that can indicate possible diversion of the spent nuclear fuel rods from the spent nuclear fuel array in the spent nuclear fuel storage pool. 15. A method of determining possible diversion of spent nuclear fuel rods located in a spent nuclear fuel storage pool in a spent nuclear fuel array in the spent nuclear fuel storage pool, wherein the spent nuclear fuel array includes spent nuclear fuel rods arranged in a fuel rods quadrant symmetric pattern and guide tubes adjacent and between the spent nuclear fuel rods, comprising the following steps:the step of providing a detector instrument insertion fixture having a slider assembly with insertion tubes arranged in a detector quadrant symmetric pattern corresponding to the fuel rods quadrant symmetric pattern, said insertion tubes containing a first detector and a second detector, wherein said first detector is a neutron detector and said second detector is a gamma detector,the step of moving said detector instrument insertion fixture into the spent nuclear fuel storage pool adjacent the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of moving said slider assembly with said insertion tubes into contact with the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of measuring radiation responses, both gammas and neutrons, of said first detector and said second detector simultaneously at a location or multiple locations within said detector quadrant symmetric pattern,the step of determining ratios of the total gamma flux and neutron flux obtained at each measurement,the step of normalizing said ratios to the maximum among them to obtain a unique profile signature, andthe step of determining wherein said unique profile signature has been perturbed in said detector quadrant symmetric pattern and determining whether spent nuclear fuel rods are missing or whether spent nuclear fuel rods have been replaced with dummy spent nuclear fuel rods using said unique profile signature in the spent nuclear fuel storage pool. 16. A method of determining possible diversion of spent nuclear fuel rods located in a spent nuclear fuel storage pool in a spent nuclear fuel array in the spent nuclear fuel storage pool, wherein the spent nuclear fuel array includes spent nuclear fuel rods arranged in a fuel rods quadrant symmetric pattern and guide tubes adjacent and between the spent nuclear fuel rods, comprising the following steps:the step of providing a detector instrument insertion fixture having a slider assembly with insertion tubes arranged in a detector quadrant symmetric pattern corresponding to the fuel rods quadrant symmetric pattern, said insertion tubes containing a first detector and a second detector, wherein said first detector is a gamma detector and said second detector is a neutron detector,the step of moving said detector instrument insertion fixture into the spent nuclear fuel storage pool adjacent the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of moving said slider assembly with said insertion tubes into contact with the spent nuclear fuel array in the spent nuclear fuel storage pool,the step of inserting said insertion tubes into said guide tubes in the spent nuclear fuel array in the spent nuclear fuel storage pool using said detector instrument insertion fixture,the step of measuring radiation responses, both gammas and neutrons, of said first detector and said second detector simultaneously at a location or multiple locations within said detector quadrant symmetric pattern,the step of determining ratios of the total gamma flux and neutron flux are obtained at each measurement,the step of normalizing said ratios to the maximum among them to obtain a unique profile signature, andthe step of determining wherein said unique profile signature has been perturbed in said detector quadrant symmetric pattern and whether there are missing spent nuclear fuel rods or whether spent nuclear fuel rods have been replaced with dummy spent nuclear fuel rods in said spent nuclear fuel storage pool. |
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claims | 1. A radiation detection system, comprising:a radiation adsorption bed;an air inlet;an air outlet; anda pump operably connected to the air inlet or air outlet configured to create a flow of ambient air through the air inlet, over the radiation adsorption bed, and out the air outlet. 2. The system of claim 1, wherein the radiation adsorption bed comprises an activated-carbon sorbent. 3. The system of claim 1, further comprising a temperature sensor positioned to measure an internal temperature of the system. 4. The system of claim 3, further comprising a heating element operably coupled to the temperature sensor and configured to heat the adsorption bed based on the internal temperature of the system. 5. The system of claim 3, further comprising a processor operatively coupled to the temperature sensor and a network connection device and configured to store the internal temperature of the system in at least one of a local storage device and a remote storage device. 6. The system of claim 1, further comprising a radiation monitor positioned to measure a radiation level in the radiation adsorption bed. 7. The system of claim 6, further comprising a heating element operably coupled to the radiation monitor and configured to heat based on the radiation level of the system. 8. The system of claim 7, wherein the heating element raises the temperature of the radiation adsorption bed to about 60° C. and about 150° C. 9. The system of claim 1, further comprising a housing encompassing the radiation adsorption bed, air inlet, air outlet, and pump. 10. The system of claim 9, wherein the housing is configured as a wearable device. |
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