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abstract | In order to realize a filter that can make a fine spectrum adjustment in a wide rage and can be miniaturized, and to realize an X-ray imaging apparatus provided with the filter, a filter of the present invention is provided with plural filter plates that can form a layer crossing X-ray and adjusting means that adjusts a combination of filter plates forming the layer by individually moving the plural filter plates so as to come in and out the X-ray passing space. The plural filter plates are formed such that the thickness of each filter plate is successively doubled with the thinnest filter plate defined as a reference. The adjusting means has a pair of moving in/out mechanisms for moving, every its half number, the plural filter plates in and out from both sides of the X-ray passing space. |
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059463643 | summary | FIELD OF THE INVENTION This invention relates to the evaluation of the densification of fissionable nuclear reactor fuel comprised at least in part of an oxide of uranium. BACKGROUND OF THE INVENTION Fissionable nuclear fuel for nuclear reactors typically comprises one of two principal chemical forms. One type consists of fissionable elements such as uranium, plutonium and thorium, and mixtures thereof, in metallic, non-oxide form. Specifically this category comprises uranium, plutonium, etc. metal and mixtures of these metals, namely alloys of such metals. The other principal type of nuclear reactor fuel consists of ceramic or non-metallic oxides of fissionable and/or fertile elements comprising uranium, plutonium or thorium, and mixtures thereof. This category of ceramic or oxide fuels is disclosed, for example, in U.S. Pat. No. 4,200,492, issued Apr. 29, 1980, and U.S. Pat. No. 4,372,817, issued Feb. 8, 1983. Uranium oxides, especially uranium dioxide, has become the standard form of fissionable fuel in commercial nuclear power plants used for the generation of electrical power. However, minor amounts of other fissionable materials such as plutonium oxide and thorium oxide, and/or neutron absorbers, sometimes referred to as "poisons", such as gadolinium oxide, are sometimes admixed with the uranium oxide in the fuel product. Uranium oxide fuel is generally produced by converting uranium hexafluoride or uranium metal to oxides of uranium. The process includes a series of chemical and physical operations, including pressure compacting uranium oxide in particulate form into handlable pellets or physically integrated bodies of suitable size and configuration, then sintering the resultant pellets or bodies of compacted particles. Sintering at high temperature coalesces the compacted particles of each pellet or body into an integrated unit of high density, and produces other desired effects such as manipulating the molecular oxygen content of the material and removal of residual undesirable impurities, e.g., fluorides. Sintering processes are amply disclosed in the art, for example U.S. Pat. No. 3,375,306, issued Mar. 26, 1968; U.S. Pat. No. 3,872,022, issued Mar. 18, 1975; U.S. Pat. No. 3,883,623, issued May 13, 1975; U.S. Pat. No. 3,923,933, issued Dec. 2, 1975; U.S. Pat. No. 3,930,787, issued Jan. 6, 1976; U.S. Pat. No. 4,052,330, issued Oct. 4, 1977; and U.S. Pat. No. 4,348,339, issued Sep. 7, 1982. Fissionable nuclear fuel materials for commercial power generating, water cooled and/or moderated reactors, commonly comprising pellets of uranium oxide, are typically enclosed within a sealed container formed of an alloy of zirconium metal, such as zircaloy -2 (U.S. Pat. No. 2,722,964), or possibly stainless steel, to provide a fuel element. The container, sometimes referred to in the nuclear field as "cladding", generally comprises a tube-like or elongated enclosure housing fuel pellets stacked therein end-on-end to the extent of about 3/4 of the length of the containers. Fissionable fuel is enclosed and sealed in such containers for service in nuclear reactors to isolate it from contact with the coolant and/or liquid moderator. This precludes either any reaction between the fuel or fission products and the coolant or moderator media, or contamination of the coolant or moderator with escaping radioactive matter from the fuel or fission products. Experience has shown that after extensive exposure to the radiation in the core of an operating nuclear reactor, typical fuel elements consisting of the fissionable fuel sealed within a metal container, are susceptible to failures due to breaching of their containers during rapid power increases. Fuel container breaching has been determined to be a result of a combination of conditions, namely stress imposed upon the metal by thermal expansion of the contained fuel, embrittlement of the metal by prolonged exposure to radiation and corrosion by the presence of accumulated fission products from the fuel enclosed therein. Studies of this deleterious phenomenon have determined that these three conditions contribute to produce such a failure of the metal fuel container, which is commonly referred to in the art as "intergranular stress corrosion cracking". First, the metal must be susceptible to stress corrosion cracking in the irradiation environment; secondly, a level of physical stress must be present; and, thirdly, there must be exposure to aggressive corrosive agents. Metal failure due to stress corrosion cracking can be mitigated or even eliminated by alleviating any one or more of these three conditions. One effective means for deterring such failures in conventional fuel elements comprising zirconium alloy containers housing uranium oxide fuel has been to include a metallurgically bonded barrier liner of unalloyed zirconium metal over the inner surface of the alloy container substrate. The unalloyed zirconium metal of the barrier liner is more resistant to irradiation embrittlement than the alloy substrate whereby it retains its initial relatively soft and plastic characteristics throughout its service life notwithstanding prolonged exposure to irradiations, etc. Localized physical stresses imposed on such a barrier lined fuel container by heat expanding fuel during rapid power increases are moderated by the plastic movement of the relatively soft unalloyed zirconium metal of the liner. Moreover, the unalloyed zirconium metal has been found to be less susceptible than alloys to the effects of corrosive fission products. That is, the unalloyed zirconium has resistance to the propagation of cracks in the presence of corrosive fission products. The effectiveness of the unalloyed zirconium barrier liner in resisting the deleterious stress corrosion cracking phenomenon due to the interaction between the fuel pellets and the container in the presence of a corrosive environment of irradiation products, is achieved by mitigating the physical stress and stress corrosion crack propagation susceptibility of the zirconium barrier layer. Effective unalloyed zirconium metal barrier linings for nuclear fuel elements comprising fuel pellets enclosed within a container are disclosed in U.S. Pat. No. 4,200,492 and No. 4,372,817. Another approach to the problem of stress corrosion cracking as a cause of failure of fuel elements when subjected to frequent and drastic power increase has been to modify the physical properties of the uranium oxide fuel with the inclusion of additives. For example, aluminum silicates, derived from clays, when dispersed throughout the uranium oxide in amounts as low as a few tenths of one percent, have been demonstrated to be effective in increasing the plasticity of fuel pellets composed thereof, whereby the thermal expansion induced physical stress attributable to the fuel pellets is reduced. The aluminum silicate may also play a role in reducing the effectiveness and availability of the chemically aggressive fission products which promote stress corrosion cracking of the cladding tubes. Aluminum silicate additives blended with uranium oxide have been found to be effective in eliminating or mitigating two of the three conditions which must be simultaneously present to produce stress corrosion failures in the metal of a fuel container. An aluminum silicate additive substantially increases the creep rate of fuel pellets comprising oxides of uranium and thereby reduces the stress imposed on the container due to thermal expansion of the fuel material. The enhanced plastic deformation and deformation rates attributable to this additive enables the modified fuel to flow into its own void volume within the interior of the fuel container, and thereby distribute the physical interaction force due to thermal expansion over a greater area. Thus high localized stresses are mitigated by increased distribution of their forces. Moreover, the aluminum silicate introduced into the fuel material reacts with fission products produced during irradiation and accordingly reduces the concentration of aggressive fission products which, in the presence of physical stresses, are a cause of cracking in the metal of the fuel containers. The effects of additives comprising aluminum silicates upon fissionable nuclear fuels, including their relative quantities, are disclosed in U.S. Pat. No. 3,679,596; No. 3,715,273; No. 3,826,754; No. 3,872,022; and No. 4,052,330. However, experience in the processing or fabrication of aluminum silicate containing ceramic fuels comprising oxides of fissionable elements employing the conventional sintering procedures and conditions used for ceramic fuel has demonstrated the occurrence of distinctive shortcomings in the resulting products. Specifically it has been found that there occurs inconsistencies in the concentrations of aluminum silicate added and in achieving the final fuel densities desired. The conventional sintering procedures and conditions commonly used in producing fuel with uranium oxides, such as disclosed in the foregoing patents, comprises employing reducing conditions to provide for an oxygen to metal ratio of the fuel material of near or at the desired stoichiometric composition of oxygen to metal ratio O/M=2.00 (UO.sub.2) during and following the sintering operation. For example, hydrogen or cracked ammonia sintering atmospheres with relatively low dew points, such as <10 degrees C., or hydrogen/carbon dioxide gas mixtures or carbon monoxide/carbon dioxide gas mixtures with their ratios proportionally adjusted to produce near the stoichiometric compositions are typically used in sintering. Reducing conditions with high sintering temperatures, such as about 1700 degrees C. or higher result in a relatively high vapor pressure of silicon monoxide (SiO) over silicon dioxide (SiO.sub.2) and aluminosilicate, amounting to as much as a few tenths of an atmosphere. See for instance "Graphical Displays of the Thermodynamics of High-Temperature Gas-Solid Reactions and Their Application to Oxidation of Metals and Evaporation of Oxides" by Lou et al, Journal of the American Ceramic Society, Vol. 68, No. 2 February 1985, pages 49-58. Due to such high SiO vapor pressures, there is considerable volatilization of the silica bearing material from a uranium oxide material such as a fissionable fuel composition containing an aluminosilicate or silica bearing phase. Such a loss of silica material presents difficulties in controlling the amount of silica containing additives present in a fuel product. Moreover, because of the high vapor pressure of SiO over the silica containing additive phase, pores or voids formed within the additive phase are stabilized and achieving the desired final density is inhibited. The disclosed contents of the foregoing U.S. Pat. namely No. 3,375,306; No. 3,679,596; No. 3,715,273; No. 3,826,754; No. 3,872,022; No. 3,883,623; No. 3,923,933; No. 3,930,787; No. 4,052,330; No. 4,348,339; No. 4,578,229; No. 4,200,492; and No. 4,372,817, which illustrate the state of the art relevant to the invention disclosed and claimed herein, are each incorporated herein by reference. SUMMARY OF THE INVENTION This invention comprises an improved method of testing nuclear fuel products comprising an oxide of uranium which may typically incorporate a silica-containing additive. The invention includes a high temperature test procedure for determining density changes wherein the atmospheric composition is regulated to inhibit losses of silica-containing additive which may be present. OBJECTS OF THE INVENTION It is a primary object of this invention to provide an improved test procedure for evaluating densification of a fissionable nuclear fuel product comprising an oxide of uranium. It is also an object of this invention to provide an improved test comprising thermal treatment of a nuclear fuel composition of an oxide of uranium and a silica-containing additive for use in the manufacture of fissionable fuel products. It is a further object of this invention to provide a test procedure for use in the manufacture of nuclear fuel comprising uranium oxide with a silica-containing additive which inhibits composition changes due to a loss of the silica-containing additive during thermal treatment. It is an additional object of this invention to provide a method for testing nuclear fuel comprising uranium oxide with an aluminum silicate additive which enables a determination of the product density changes during service within a nuclear reactor. It is a still further object of this invention to provide a means of impeding loss of SiO and in turn unwanted compositional changes during density evaluations which include thermal treatment. DETAILED DESCRIPTION OF THE INVENTION This invention deals with nuclear fuel products produced from fissionable materials comprising oxides of uranium which may include a silica-containing additive such as disclosed in the above patents. The fissionable material, in addition to the uranium oxide and silica-containing additive (if present), can also include oxides of plutonium or thorium, neutron absorbers or "poisons" such as gadolinia, and combinations thereof, among other ingredients disclosed in the above cited prior art. The oxides of uranium and other fissionable ceramics are preferably of an oxygen to metal ratio (O/M) of approximately 2.00, namely substantially composed of uranium dioxide (UO.sub.2). The silica-containing additives, likewise include those disclosed, and their amounts, as given in the above-cited patents. Specific silica containing additives include silicon dioxide (SiO.sub.2), aluminum silicates (Al.sub.2 O.sub.3.SiO.sub.2) and derivatives thereof, including natural clays such as mullite (3Al.sub.2 O.sub.3.2SiO.sub.2) pyrophillites (Al.sub.2 O.sub.3.4SiO.sub.2), kaolinite (Al.sub.2 O.sub.3 (Si.sub.2 O.sub.3)).(OH).sub.4), andalusite (Al.sub.2 SiO.sub.3), sillimanite (Al.sub.2 SiO.sub.5), and cyanite (Al.sub.2 SiO.sub.5), for example. Mixtures of alumina powder and silica powder may also be employed. Alternatively, it is possible to introduce each of the silicon and aluminum as a compound which decomposes to silica and alumina under the conditions of sintering. For example, the aluminum, or at least a portion of it, may be added as an organoaluminum compound, such as for example aluminum bistearate, diethylaluminum malonate or triphenyl aluminum. The aluminum compound, especially the bistearate, would act as a pressing die lubricant, and leave alumina when the hydrocarbon portion is volatilized. An organosilicon compound may be used for the silica addition, such as for example a volatile silicon compound that will vaporize early in the sintering process. Examples include silicobenzoic acid, triethylphenylsilicane, ethyltriphenylsilicane and methyltriphenyl silicane. The organosilicon compound would produce the fugitive silicon which would be converted to silica in the sintering furnace, and would act as a pore former to control the density and structure of the sintered pellets. Amounts of the silica-containing additives comprise, for example, about 0.025 percent up to about 1.0 percent by weight of the overall fuel material. With the sintering conditions commonly employed in the manufacture of uranium oxide fuel, the vapor pressure of SiO is strongly dependent upon temperature and oxygen free energy. For example, at 1700 degrees C., the SiO vapor pressure can range from approximately 10.sup.-6 (0.000001) to 10.sup.-1 (0.10) atmospheres, note "Review-Graphic Displays of the Thermodynamics of High Temperature Gas-Solid Reactions and Their Application to Oxidation of Metals and Evaporation of Oxides", by Lou et al, supra. At the typical sintering conditions used for urania-based nuclear fuels, about 1700-1800 degrees C., the vapor pressure of SiO is near 10.sup.-2 (0.01) atmospheres. Under such conditions, there can occur a considerable loss of any silica bearing material. As reported in detail by R. O. Meyer, in "The Analysis of Fuel Densification", U.S. Nuclear Regulatory Commission, Report NUREG-0085, July 1976, due to the occurrence of fuel containers flattening or collapsing in reactor service, an effort was made to evaluate an observed phenomenon of fuel densification, or shrinkage. This undertaking resulted in a test procedure for nuclear fuel to evaluate the potential of a given fuel product for densification or shrinkage during operation in-reactor. The densification evaluation test procedure which evolved for simulating in reactor densification, and as described in the above Meyer's article, briefly comprises heat treating produced fuel specimens (pellet samples of fuel products) at a temperature of 1700 degrees C. (3092 degrees F.) for 24 hours in dry hydrogen, or low dew point (for example <10.degree. C.) atmosphere. Efforts in undertaking to apply this standardized densification test procedure to aluminosilicate containing ceramic fuels have demonstrated that there are unique problems which generally prevent the obtaining of meaningful test results. The test specimens commonly exhibited significant losses in weight, and occasionally actually increases in volume during the procedure, rather than showing the expected densification. The standard thermal simulation test procedure and conditions for a ceramic fuel comprising uranium oxide based fuel products comprised providing relatively reducing conditions to maintain the oxygen to uranium ratio (O/U) of the fuel material at or near the stoichiometric composition of uranium dioxide (O/U=2.00) during the testing. It has been determined that under the high temperature (about 1700 degrees C.) reducing conditions, the vapor pressure of silicon oxide (SiO) over silicon dioxide (SiO.sub.2) and alumino-silicates is quite high, amounting to as much as several tenths of an atmosphere. Due to this high vapor pressure there occurs considerable volatilization of the silica bearing material from the uranium oxide fuel containing an aluminosilicate or silica phase. This volatilization phenomenon causes the observed weight loss while carrying out the densification test procedure with an aluminosilicate containing fuel. Moreover, due to the high vapor pressure of the silica oxide (SiO) over the additive phase, pores within the additive phase are stabilized and may even grow, thereby producing the observed swelling of fuel during the test and confounding the test results. The vapor pressure of silica oxide (SiO) has been determined to be significantly dependent upon ambient temperature and oxygen free energy. For instance, at 1700 degrees C. the silica oxide (SiO) vapor pressure can range from approximately 0.000001 to 0.1 atmospheres as noted above. At the thermal simulation test conditions used for assessing the resistance of uranium oxide based nuclear fuel to in-reactor service conditions, the vapor pressure of silica oxide (SiO) is near the highest possible value at 1700 degrees C. In accordance with this invention, the subject thermal simulation testing of uranium oxide fuel containing an aluminosilicate or silica phase for ascertaining its resistance to densification when in reactor service, is carried out while increasing the oxygen free energy a significant degree, namely several kilocalories/mole, thereby reducing the equilibrium vapor pressure of SiO by several orders of magnitude. For instance, when the partial molar free energy of oxygen during the thermal simulating test for ascertaining resistance of the fuel to densification during in-reactor service, is increased to about -80 kilocalories per mole or more, the silica oxide (SiO) vapor pressure while performing the test at 1700 degrees C. is decreased from nearly 0.1 atmospheres down to only about 0.0001 atmospheres or less. The rate of evaporation of the silica oxide (SiO) from the fuel composition, in turn is similarly reduced by about three orders of magnitude or more, thereby effectively eliminating the weight loss deficiency in performing the test. Additionally, the low SiO vapor pressure eliminates stabilization and growth of pores within the pellet or body of compacted and sintered particulate uranium oxide fuel containing a silica phase due to internal pressurization. Thus the confounding effect of testing aluminosilicate containing uranium oxide based fuel material is overcome. The preferred conditions for the practice of this invention comprise carrying out the thermal simulation test procedure for uranium oxide based nuclear fuel materials containing silicon dioxide or aluminum silicate additives in an atmosphere which produces a low SiO vapor pressure by providing and maintaining the partial molar free energy of oxygen therein of greater than -90 kilocalories per mole. Oxygen partial molar free energy can be regulated by manipulating the gas composition of the testing atmosphere such as by applying specific gases and or by proportioning the ratios of mixtures of gases. For example, the testing atmosphere conditions can be achieved through the application of wet hydrogen, wet cracked ammonia, mixtures of carbon monoxide/carbon dioxide gases and mixtures of hydrogen/carbon dioxide gases in appropriate ratios for the selected test temperature and gas atmosphere. Preferred testing temperatures for the practice of this invention comprise a range of from about 1700 degrees C. (3092 degrees F.) up to about 2200 degrees C. (3992 degrees F.). |
061954063 | claims | 1. A pressurizer of a nuclear power plant comprising: a casing having a wall and a lower region; a volume-compensation line ending in said casing; and a spray line only intended for cooling, said spray line guided through said wall in said lower region, extended upward inside said casing and ending in said casing at a geodetically highest point of said spray line. 2. The pressurizer of a nuclear power plant according to claim 1, wherein said spray line has at least one spray nozzle in the vicinity of said highest point. 3. The pressurizer of a nuclear power plant according to claim 2, wherein said spray line has a portion directed upward at an angle, said portion having said at least one spray nozzle. 4. The pressurizer of a nuclear power plant according to claim 1, wherein said lower region of said casing is a lower cylindrical part, and said spray line is guided through said wall in said lower cylindrical part at an oblique angle relative to said wall. 5. The pressurizer of a nuclear power plant according to claim 1, wherein said lower region of said casing is a lower cylindrical part, and said spray line is guided through said wall in said lower cylindrical part at right angles to said wall. 6. The pressurizer of a nuclear power plant according to claim 1, wherein said casing has a bottom and a dome-like part closing off said bottom, and said spray line is guided through said wall in said dome-like part at right angles to said wall. 7. The pressurizer of a nuclear power plant according to claim 1, wherein said spray line is guided through said wall at a location forming a fixed point. |
055286399 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 is a typical BWR power/flow operating map showing a conventional protection system having the alarm setpoint a distance A above the operating point 1 and the scram setpoint a distance B above the operating point 1, both setpoints being above the maximum operating line. After startup, the permissible operating range for a BWR is above the cavitation region, below the maximum operating line and bounded by the minimum normal flow line and the maximum normal flow line. In conventional protection systems, when the BWR is operating within the operating zone, an unplanned transient that does not increase the power level above the maximum operating line will not be detected by the setpoints and reactor trip will not occur. The invention overcomes this problem by providing safety system setpoints (at which transient mitigation action is initiated) which are adjusted so that they are much closer to the operating power level (for example, point 1) than conventional protection designs. This principle can be applied as needed throughout the entire normal power/flow operating range shown in FIG. 1. FIG. 2 is an example of a BWR power/flow operating map showing the enhanced protection provided by the invention. If the reactor is operating at 100% power or along the maximum operating line, the setpoints provided by the invention will be automatically adjusted to essentially the same position as in conventional protection systems (a distance of either A or B above operating point 1 in FIG. 1). However, if the reactor is operating at a partial power condition (such as point 1 in FIG. 2), the invention provides alarm and scram setpoints that are closer to that point. In FIG. 2, A1 and A2 represent the adjusted margins between the operating point and two alarm setpoints and B represents the adjusted margin between the operating point and a scram setpoint. The tracking logic in accordance with the invention controls the adjustment of the alarm and scram setpoints so that they are set the desired amount (A1, A2 and B) above any operating condition within the BWR's range of operation. Typical signals and functions included in the invention are shown in FIG. 4. The new portions of logic added by the invention are demarcated from conventional design elements by a dashed line 10. This simplified diagram is-intended to illustrate the essential principles of the invention. It does not show the redundancy necessary for reactor protection functions, nor is it to be construed as the only manner in which the functional logic of the invention can be implemented. Referring to FIG. 4, the invention's tracking scram setpoint logic 12 and tracking alarm setpoint logic 14 (only one alarm function is shown for simplicity) maintain the desired trip margin during planned power increases by automatically increasing the scram and alarm setpoints. These setpoints are respectively used by the scram trip unit 16 and alarm trip unit 18 to monitor the STP signal output from filter 20. Reactor scram can be initiated by the output of a scram signal from either scram trip unit 16 or high power trip unit 22. Planned power changes are identified by a permissive input signal which may be generated manually by the reactor operator or in association with normal methods of increasing power (e.g., control rod withdrawals or recirculation flow setpoint increases). However, when an unplanned power increase occurs, the tracking logic will not increase the setpoints except as controlled, thereby providing enhanced protection. The setpoint adjustment will also automatically track any reactor maneuver which significantly reduces the power level. In this way the protection setpoints are re-established near the new, final operating point. The upper and lower values of the setpoints may also be limited to bound the function of the invention to a desired operating range. The tracking logic of the invention may use one or more alarms (A1, A2) in conjunction with the STP scram signal (B) (shown in FIG. 2). The use of alarm signals to perform active functions to avoid full reactor trip (or scram) is another important attribute of the invention. In addition to alerting the operator, various actions may be initiated at the alarm setpoint(s) (A1, A2) to stop the power increase without imposing the operational penalties associated with total shutdown of the reactor caused by scram. Such actions include, but are not limited to, blocking of control rod withdrawal, reduction of reactor recirculation flow and insertion of selected control rods. An additional feature of the invention is the option to include supplementary adjustment of the high-power trip setpoints based on signals from other reactor parameters. For example, the setpoints may be adjusted in dependence on reactor pressure, reactor recirculation flow or feedwater temperature. The reactor protection system in accordance with the invention also includes the ability to use either the filtered STP signal and/or the direct neutron flux signal (i.e., the "power range monitor signal" in FIG. 4) as the input to the tracking scram setpoint logic 12. The setpoints used with a direct neutron flux signal must be set higher than those used with the STP filter 20 method to avoid inadvertent actuation. FIG. 3 shows an example of an application of the invention with tracking setpoints supplemented by a recirculation flow signal. If the reactor is operating at full power or along the maximum operating line, it is similar in many respects to the example shown in FIG. 2, with the maximum expected setpoints adjusted to be essentially equal to the setpoints of conventional protection systems. The setpoints are also adjusted to be the desired margin above any partial power operating point (for example, operating point 1 in FIG. 3). The unique aspect of this application is that the scram and alarm setpoints are also varied automatically with changes in reactor recirculation flow (the setpoints have a flow-referenced slope above point 1 in FIG. 3). Also shown in FIG. 3 is an example of the alternative to use a direct neutron flux signal in conjunction with the tracking setpoint logic (set above the STP setpoints at C). The amount of the variation with flow (the slope of the setpoint lines above point 1 in FIG. 3) can be chosen to optimize performance of the invention during reactor flow and power maneuvers. The variation of the setpoints with supplemental reactor parameters (e.g., recirculation flow in this example) may also be limited in magnitude and/or direction of change to optimize the effectiveness of the application of the invention. If the flow-referenced option is used, as shown in FIG. 3, the setpoints will also be automatically increased if the power increase is caused by an unplanned reactor recirculation flow increase. However, the setpoint increase will be a controlled amount according to the slope of the flow-dependent setpoint variation. The invention therefore provides enhanced reactor protection by adjusting the trip setpoints so that they remain close to the operating point anywhere in the power/flow operating range of the reactor. In conjunction with this closer safety trip (scram) protection, the invention provides alarms that are simultaneously adjusted so that automatic actions can also be initiated to avoid full shutdown of the unit during transient events. The setpoints automatically track power decreases, but increases of the setpoints are restricted so that they provide enhanced protection for all unplanned transients that increase reactor power. The tracking protection of the invention responds favorably to simulated reactor transients, including the postulated, slow events. A few transient examples are presented hereinbelow to demonstrate the performance of the present invention. EXAMPLE 1 Temperature Transient, Basic Invention One type of event that can occur in a BWR is a change in the temperature of the coolant flow being supplied to the reactor core. One way that this can happen is if a portion of the feedwater heaters fail to operate properly. FIGS. 5A and 5B show the calculated response of the reactor and the enhanced protection logic over time to this type of an event. The initial power is 70% and the reactor is assumed to be operating with maximum normal core flow. This operating condition is a significant amount below the conventional scram setpoint (shown in FIG. 5B). FIGS. 5A and 5B show that as the cooler water reaches the reactor, the power gradually increases. In this case, the STP signal increases almost up to the tracking scram setpoint provided by the invention. In FIG. 5B, the margins to the tracking alarm and scram setpoints are shown as the event progresses. In this example, only one alarm was simulated, and no scram avoidance actions were initiated when the alarm was reached (near 60 seconds, well ahead of when the scram setpoint is approached). The transient simulated in FIGS. 5A and 5B is equal to the maximum change in feedwater temperature currently allowed (100.degree. F.). Any larger change in temperature is unlikely. But should it occur, it would reach the scram setpoint provided by the invention. Therefore, acceptable reactor fuel protection is assured by the reactor protection system of the present invention. In contrast, conventional systems would not provide such protection if the same event were to occur because the conventional STP setpoint is well above the power transient. Therefore, manual operator actions would be required under conventional systems to provide protection. The performance shown in this example applies primarily to the basic invention. However, it also applies to the flow-referenced logic option if the reactor core flow remains constant during the event (manual flow control). Response in automatic flow control with the flow-referenced option is provided in the next example. EXAMPLE 2 Temperature Transient, Flow-Referenced Tracking Option In this example, an unplanned temperature transient similar to the one described in Example 1 is postulated to occur, but the reactor is assumed to be operating at full power in automatic flow control mode. The purpose of the automatic flow control is to hold reactor power at the initial power level setpoint. In this control mode, the reactor recirculation flow is automatically reduced during this event to counteract the power increasing effects of the transient. FIGS. 6A, 6B and 6C show a typical response to this type of event. These figures show that as the simulated temperature change tries to increase the reactor power, the automatic controls decrease the core recirculation flow so that power remains essentially constant. FIGS. 6A and 6B show the response of key reactor parameters versus time. As in Example 1, the currently limiting magnitude of the temperature change has been simulated. The transient settles to a final operating condition without the need for any protection. Since the controlled power level is supported, however, by less core coolant flow, it is approaching a condition in which insufficient cooling may be available to the reactor fuel. FIG. 6C shows how the reactor operating point moves along at constant power, but decreasing core flow characteristic during the simulated event. The invention with the flow-referenced option reduces the tracking setpoints as recirculation flow is reduced, so that by the end of this case, the scram setpoint is just above the final operating point. Tracking alarm actions are ignored in this case. Any larger temperature transient would initiate the new protection. The existing flow-referenced scram setpoint is also shown. It follows the characteristic shown in FIG. 2, and is further away from the operating condition. If the event had been simulated at lower initial power (e.g., 70% as in Example 1), the difference between the operating point and the conventional scram setpoint would be larger, while the setpoint provided by the invention will remain close to the operating point. EXAMPLE 3 Core Flow and Power Increase One common reactor maneuver that must be accommodated without reactor trip is the normal increase of power using the reactor core flow control system. FIGS. 7A, 7B and 7C demonstrate how the invention is able to accommodate this type of maneuver. In this situation, the operators will have planned and prepared for the power increase, and the permissive logic of the invention is activated at the start of the increase. The responses of the reactor and the tracking logic of the invention are shown in FIGS. 7A and 7B. Core flow and power are increased gradually in this ramp-like maneuver. The tracking setpoints of the invention increase with the reactor power. Margin is maintained, as required, between the STP signal and the alarm and scram setpoints. FIGS. 7A and 7B shows responses of key reactor parameters versus time. FIG. 7C shows the tracking action of the setpoint logic in accordance with the invention. The trip avoidance margins for the alarms (two in this example) and the scram are shown at the bottom of FIG. 7B to stay almost equal to the initial margin throughout the maneuver. The preferred embodiments have been disclosed for the purpose of illustration only. Variations and modifications of those embodiments will be readily apparent to engineers of ordinary skill in the art of boiling water reactor protection systems. All such variations and modifications are intended to be encompassed by the claims appended hereto. |
abstract | In a scintillator panel, a glass substrate with the thickness of not more than 150 μm serves as a support body, thereby achieving excellent radiotransparency and flexibility. Furthermore, in this scintillator panel, an organic resin layer is formed so as to cover the entire surface of the glass substrate. This reinforces the glass substrate, whereby the edge part thereof can be prevented from chipping or cracking. Furthermore, stray light can be prevented from entering a side face of the glass substrate and, since the organic resin layer is formed on the entire surface, it becomes feasible to suppress warping of the glass substrate due to internal stress after formation of a scintillator layer. |
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053348435 | abstract | A scintillator screen for an X-ray system includes a substrate of low-Z material and bodies of a high-Z material embedded within the substrate. By preselecting the size of the bodies embedded within the substrate, the spacial separation of the bodies and the thickness of the screen, the sensitivity of the screen to X-rays within a predetermined energy range can be predicted. |
049820982 | description | DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS Now, the present invention will be described more specifically below with reference to working examples. EXAMPLE 1 An intensifying screen (354 mm.times.354 mm in area) for X-ray radiography of the chest was produced by the following procedure, using calcium tungstate as phosphor and fixing the magnitudes of speed in the pattern of FIG. 1 so that when the magnitude of speed in the regions A' and B' was taken as 100, that of speed in the regions H and I would fall at 40. Preparation of light-absorbing layer FIG. 7 is a cross section illustrating the manner of absorbing manufacture of a photogravure plate used for the formation of a light-absorbing layer. With reference to this diagram, 30 stands for an intensifying screen used as a planar light source. A black paper pattern 32 provided with an opening 34 of the shape conforming to a desired pattern of speed correction and intended to intercept the light emanating from the planar light source was placed in close contact with one surface of the intensifying screen 30. This black paper pattern 32 and an X-ray film 36 disposed at a distance of about 10 mm downward from the lower side of the paper pattern 32 were enclosed in a cassette 38 and exposed to a beam of X-rays as indicated by the arrow z in the diagram. Then, the X-ray film was exposed to the scattered light from the planar light source. Consequently, there was obtained a photogravure plate having density smoothly and continuously varied across each of the borderline between the part corresponding to the black paper pattern 32 and the remaining part. In the present working example, two black paper patterns were prepared which contained openings shaped as illustated respectively in FIG. 8 and FIG. 9 as desired patterns of speed correction. An X-ray film was exposed to light first through the black paper pattern 32a possessing the opening 34a shown in FIG. 8 (as hatched) and subsequently through the black paper pattern 32b possessing the openings 34b shown in FIG. 9 (as hatched), to produce a photogravure plate. Then, a printing plate was produced from the photogravure plate. A yellow pigment capable of absorbing the light emitted from calcium tungstate was applied on a film of polyester 10 .mu.m in thickness by gravure printing, with the aid of the photogravure plate. Thus, a photo-absorbing layer was formed. Preparation of intensifying screen A slurry was prepared by mixing 20% by weight of calcium tungstate as phosphor with 2% by weight of polyvinyl butyral and 78% by weight of butyl acetate as a binder. As shown in FIG. 5 second layer 14b of phosphor about 75 .mu.m in thickness was formed by applying this phosphor slurry on a protective layer 18 made of polyester film about 10 .mu.m in thickness and drying the applied layer of the slurry. Subsequently, the light-absorbing layer 16 formed by the aforementioned procedure was superposed on the second layer 14b of phosphor. On this layer 16 a first layer 14a of phosphor 75 .mu.m in thickness was formed by applying a phosphor slurry of the same composition as mentioned above on the second layer 14b of phosphor and drying the applied layer of the slurry. Thereafter, a substrate 12 made of polyester film 250 .mu.m in thickness was attached fast to the upper surface of the aforementioned first layer 14a of phosphor, to produce an intensifying screen 10 for X-ray radiography of the chest. The intensifying screen thus obtained was superposed on an X-ray film and enclosed in a cassette. This X-ray film was exposed to a beam of X-ray film was tested for density distribution with a densitometer. When the X-ray film was examined with respect to speed distribution in the intensifying screen, this intensifying screen was found to possess a speed distribution as shown in FIG. 10. When the X-ray film was tested for change in speed along the line X--X in the diagram of FIG. 10, the intensifying screen was found to possess such a density distribution as shown in FIG. 11. As clearly noted from this diagram, the speed was smoothly and continuously changed across the borderlines between the regions of speed and within the regions speed. This fact clearly indicates that no visible line patterns were found in the produced X-ray radiograph. Then, the X-ray radiography of the chest was conducted on 50 subjects, using the intensifying screen possessing the speed distribution mentioned above. Consequently, in virtually all the cases, the lungfield, the trachea and the bronchus, and even the part of the bronchus overlapping the hilum of the left lung, were clearly radiographed in highly satisfactory contrast. The intensifying screen of the present case for the Xray radiography of the chest possessed a substantially continuous change in speed as shown in FIG. 10 because it was formed by using as the photogravure plate an X-ray radiograph having the light-absorbing layer blurred with the scattered beam of light from the planar light source. As a result, the pertinent internal organs could be radiographed clearly without entailing the occurrence of drawbacks detrimental to the diagnostic examination of the part such as line patterns in the image originating in the borderlines of change in density in the light-absorbing layer. Thus, it is evident that when the intensifying screen for the X-ray radiography of the chest possesses the speed distribution illustrated in FIG. 10 is highly effective in the diagnostic examination of the chest. EXAMPLE 2 A light-absorbing layer was prepared by the following procedure in the place of the light-absorbing layer of Example 1. Preparation of light-absorbing layer First, an X-ray radiograph of the chest of a person of standard body type was prepared and it was radiographed with the focal point moved to a certain distance. Then, the photograph of the chest taken at the aforementioned distance from the focal point was again radiographed with the focal point again moved to a certain distance. When a radiograph produced after repeating this procedure was found to show the minute details and the peripheries of bone in a perfectly blurred state, it was used as a photogravure plate. Then a printing plate was produced from the photogravure plate and a light-absorbing layer was formed using the printing plate by following the procedure of Example 1. Preparation of intensifying screen Subsequently, an intensifying screen for use in the X-ray radiography of the chest was produced using the lightabsorbing layer by following the procedure of Example 1. When the chest was actually X-ray radiographed by using the intensifying screen obtained as described above, the internal organs in the chest were clearly radiographed, indicating that the intensifying screen was as effective in producing an X-ray radiograph as the intensifying screen of Example 1. EXAMPLE 3 An intensifying screen for X-ray radiography of the head (300 mm.times.250 mm) was prepared by the following procedure, using calcium tungstate as a phosphor and fixing the speed distribution such that the speed in the region L would fall at 40 where the speed in the region J was taken as 100 in the diagram of FIG. 2. Preparation of light-absorbing layer A black light-shielding plate provided in the central part thereof with a substantially elliptical opening measuring 150 mm in major diameter and 100 mm in minor diameter. An X-ray film was disposed at a distance of about 10 mm downward from the light-shielding plate and was exposed to light in the same manner as in Example 1. A printing plate was produced by using this X-ray film as a photogravure plate. Then, on a polyester film 10 .mu.m in thickness, a light-absorbing layer was formed with yellow pigment by following the procedure of Example 1. Preparation of intensifying screen Similarly to the intensifying screen of the construction illustrated in FIG. 5, the same calcium tungstate-containing slurry as used in Example 1 was applied on a protective film 18 of polyester about 10 .mu.m in thickness and the applied layer of the slurry was dried, to give rise to a second layer 14b of phosphor about 50 .mu.m in thickness. Then, the light-absorbing layer 16 formed by the method described above was superposed on the second layer 14b of phosphor. Further the phosphor slurry of the aformentioned percentage composition was applied on the light-absorbing layer 16 and the applied layer of the slurry was dried, to give rise to a first layer 14a of phosphor 50 um in thickness. Thereafter, a substrate 12 of polyester film 250 .mu.m in thickness was attached fast to the upper side of the layer 14a of phosphor, to complete an intensifying screen 10. The intensifying screen thus obtained was placed on top of an X-ray film, enclosed in a cassette, exposed to a beam of X-rays, and tested for speed distribution in the same manner as in Example 1. The produced X-ray radiograph was found to possess a speed distribution illustrated in FIG. 12. As concerns the continuity of the change in speed, the X-ray radiograph showed no discernible line pattern, indicating that the speed was changed with sufficient continuity. Then, 50 persons were subjected to clinical test by the "Mr. Towne's method" of the head using the intensifying screen possessing this speed distribution. In virtually all the X-ray radiographs produced in the test, not merely the backbone and the occipital bone but also the portion near the scalp were clearly radiographed with highly satisfactory contrast. The intensifying screen obtained in this working example for use in X-ray radiography of the head possessed a substantially continuous change in speed as illustrated in FIG. 12 and, therefore, enabled the head to be clearly radiographed even to the peripheral part such as the portion near the scalp. It did not give rise to any such detriment to the diagnostic examination as line patterns which would possibly be produced because of the borderlines in change of density in the light-absorbing layer. The intensifying screen for the X-ray radiography of the head possessing the speed distribution illustrated in FIG. 12, as obvious from the clinical results mentioned above, permits acquisition of such information concerning the portion near the scalp as has never been utilized for diagnostic examination. Thus, it is highly effective in the diagnostic examination of the head. EXAMPLE 4 An intensifying screen (300 mm.times.200 mm) for X-ray radiography of the upper and lower jaws and the peripheries thereof was prepared by the following procedure, using calcium tungstate as a phosphor and fixing the speed distribution so that the speed in the region N would fall at 65 where that of the region M was taken as 100 in FIG. 3. Preparation of light-absorbing layer A black light-shielding plate provided in the central part thereof with an opening of the shape to give a shadow as illustrated in FIG. 3 was prepared. An X-ray film was disposed at a distance of about 10 mm downward from the light-shielding plate and exposed to light in the same manner as in Example 1. A printing plate was produced by using this X-ray film as a photogravure plate. Then, a light-absorbing layer was formed on a film of polyester 10 .mu.m in thickness by depositing carbon black by following the procedure of Example 1. In this light-absorbing layer, the portions for starting speed change toward the portion corresponding to the strip of region M of high speed were located each at a distance of about 30 mm to the left and right from the center. The light-absorbing layer possessed a density change corresponding to Type 2 illustrated in FIG. 4. Preparation of intensifying screen Similarly to the intensifying screen of the construction shown in FIG. 5, the same calcium tungstate-containing slurry as used in Example 1 was applied on a protective layer 18 of polyester about 10 .mu.m in thickness and the applied layer of the slurry was dried, to give rise to a second layer 14b of phosphor about 50 .mu.m in thickness. The light absorbing layer 16 formed by the method described above was superposed on the second layer 14b of phosphor. Then, the phosphor slurry of the aforementioned composition was applied on the layer 16 and the applied layer of the slurry was dried, to give rise to a first layer 14a of phosphor 100 .mu.m in thickness. Thereafter, a substrate 12 of polyester film incorporating therein titanium oxide or carbon black and having a thickness of 250 .mu.m was attached fast on the first layer 14a of phosphor, to give rise to an intensifying screen 10 for X-ray radiography of the upper and lower jaws and the peripheries thereof. Then, 50 persons were subjected to diagnostic examination by the X-ray radiography of the upper and lower jaws and the peripheries thereof, using the intensifying screen obtained as described above. In all the X-ray radiographs thus obtained, the parts were radiographed with substantially uniform density without suffering the shadow of the backbone to impair the distribution of radiographic density. In accordance with present working example, since the light-absorbing layer was incorporated in the layer of phosphor in such a manner as to form a strip of region of high speed at the position practically corresponding to the backbone, the amount of light emitted in the portion corresponding to the backbone was increased enough to permit production of a highly desirable X-ray radiograph of the upper and lower jaws and the peripheries thereof without giving rise to a portion of uneven-density due to the difference in X-ray absorption between the backbone and the other parts. In the category of dentistry and surgery specializing in oral cavity, therefore, this invention enables diagnostic examination of the pertinent parts of the human body to be effected accurately with one X-ray radiography. |
048572616 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings in detail, FIG. 1 shows a partial side elevation view of a typical nuclear reactor pressure vessel 10. The pressure vessel 10 includes the closure head 13 and an upwardly extending cooling shroud 16 thereon. The cooling shroud 16 has a plurality of openings 19, which can be either access doors or viewing windows, located around its circumference. On top of the cooling shroud 16 is a missile shield 22 for protecting the control rod drive mechanisms and instrumentation ports (not shown). Mounted on the cooling shroud 16 so as to view the reactor vessel head area is the reactor vessel head area monitoring system 25 of the present invention. See FIGS. 2 and 3. Preferably the monitoring system 25 is attached to the cooling shroud 16 around its circumference at areas adjacent the openings 19. The monitoring system 25 includes a video camera 28 attached to the shroud 16 Preferably the camera 28 is attached to an access door 29, which may be connected to the shroud 16 by means of a hinge 30, on the exterior surface thereof The door 29 has an aperture 31 through which the camera 28 receives video images of the interior of the cooling shroud 16 In this manner, maintenance on the video camera 28 or other components is facilitated A right angled lens 32 is attached to the video camera 28 such that the camera receives video images from the interior of the shroud area, or the reactor vessel head area. Since the camera 28 is preferably vertically oriented on the shroud 16 so as to occupy as little room as possible, a right angled lens 32, or other suitable lens having a wide angle field of view, will provide maximum viewing area. Attached to the shroud opening 19 adjacent to the video camera 28 is a light source 34, preferably comprising two halogen lamps which are variable in intensity, for reasons which will be more fully explained hereinafter. If need be, the camera 28 may be enclosed within a case 37 constructed of neutron shielding material, should the camera 28 need to be placed in areas of high neutron activity. Here, too, the right angled, wide angled lens 32 provides an important advantage in that the most sensitive element of the camera, its electronics 40, is protected from bombardment by neutrons along the neutron transport path, as indicated by dashed line 43 in FIG. 2 Alternatively, the camera 28 can be one which has been radiation hardened, such as the ETV-1250 TV Camera, manufactured by the Imaging and Sensing Technology Corporation. Preferably, the light source 34 comprises two halogen lamps 46 and 47; one lamp 46 is a primary light source and the other lamp 47 is a backup should the first one fail. Since the interior area of the reactor vessel cooling shroud 16 is quite dark, the halogen lamps 46, 47 will provide adequate lighting for the remote monitoring system 25. Preferably the power source for the camera 28 and the light source 34 are provided by a single cable 50. Also, the video camera 28 and lamps 46, 47 are releasably secured to the cooling shroud 16 so that they can be quickly removed should they need to be serviced or replaced. Since the ambient temperature around the reactor vessel 10 is relatively high, a means for supplying cooling air around the video camera 28 is provided. Cooling air, preferably plant air under pressure, is circulated around the camera 28, even if it is enclosed within the neutron shielding case 37, through an inlet 53 near an upper end 56 of the video camera 28. The air circulates around the camera 28, including its sensitive electronics 40, and outward through an outlet 59 near a bottom end 62. Preferably a plurality of such video cameras 28 and light source 34 systems are provided around the circumference of the cooling shroud 16 to provide for complete viewing of the interior of the shroud area, as well as all of the instrumentation ports and control rod drive mechanisms projecting through the vessel head 13. For most plants, three such cameras 28 are sufficient to provide complete viewing, although if need be more cameras may be provided. The cable 50 for the camera 28 and light source 34 is connected to a switching and multiplexer unit, designated as at 66, by a quick disconnect plug 67. The multiplexer unit 66 has a plurality of ports 68 to which additional monitoring systems 25 and/or lamps 46, 47 can be connected. The switching and multiplexer unit 66 allows an operator to select which video camera 28 is to be operating to inspect a particular area of the reactor vessel head. Also the particular lamp 46, 47 which is to be illuminated can be individually chosen for reasons which will become readily apparent. While a plurality of cables 50 connect to the switching and multiplexer unit 66, one for each video camera 28 and light source 34, preferably only one power cable 69 passes through the containment wall 72 of the nuclear power plant connecting to a control unit 75 disposed within a control room of the plant, forming a single penetration 76 in the containment wall 72. Preferably, most of the elements of the system are situated within the control room and not in the containment area. This way, should any of the elements need servicing, it is not necessary for personnel to enter into the irradiated environment of the containment building. Also, the number of penetrations 76 required in the containment wall 72 are to be kept at a minimum. Included with the control unit 75 is a video display 78, as well as a means 81 for retrievably storing the video images received through the video cameras 28. Preferably the storage means 81 is a video cassette recorder, or alternatively a camera to provide a snapshot of the reactor vessel head area. In this manner, should an instrumentation port or control rod drive mechanism be detected to have a leak, its condition over time can be monitored by viewing and comparing the videotapes or photographs to detect any change over the operating period of the reactor vessel 10. The control unit allows the operator to not only select which video camera 28 is operable, but also the lamp 46, 47 for each camera which is to be illuminated and its intensity as well. While only one camera 28 may be operable to send video images to the display unit 78, more than one light source 34 can be illuminated so as to allow for a unique lighting characteristic to be provided for within the interior area of the cooling shroud 16. To adequately and extensively view the interior area of the cooling shroud 16, the reactor vessel head area monitoring system 25 is preferably operated in the following manner: When it is desired to view the vessel head area, and just prior to activating the desired video camera 28, cooling air is circulated around the body of the camera 28. This will lower the temperature around the camera and dissipate heat from the electronics 40 as a particular camera 28 is operated. After a predetermined period, the video camera 28 is activated as well as one or more of the halogen lamps 46, 47. As the operator views the interior of the cooling shroud 16, a unique feature of the device allows the operator to view images similar to a three dimensional picture. The operator can selectively choose not only which lamp is illuminated, but its intensity level as well. By doing so, different types of shadowing patterns will be produced around the instrumentation ports and control drive mechanism housings. Should boric acid crystals be present on the reactor vessel head 13, these would also produce differing shadow patterns. By selectively choosing the light source 34 which is to be illuminated and its intensity, a relative indication of the size of any leak present can be determined by the amount of boric acid which may have accumulated on the reactor vessel head 13 around a particular port. When the operator has effectively viewed a particular area and recorded the video images received, the operator can selectively choose the next camera 28 and light source 34 to be energized. The above method is then performed for all of the monitoring systems 25 provided on the shroud 16 to view the entire reactor vessel head area. After the entire area has been inspected, the operator will have determined if any leaks are present and if so, whether they require immediate attention. If the amount of leakage is below that requiring immediate action, the operator will nonetheless have been alerted to the condition present. This way any such leak that is determined can be more closely monitored to determine if conditions have changed from time to time. For example, if a plant operator would normally perform such inspection for a particular plant on the order of once a month, if a leak is detected, monitoring can be performed much sooner. That is, the monitoring can be performed either weekly or daily so that should the size of the leak change, necessary steps can be taken to perform the necessary maintenance procedure. Also, the plant operator can prepare to take measures to repair the leak at the next scheduled maintenance outage, such as for refueling, for the plant. In this way, it is not necessary to shut down the plant at an unscheduled period of time, thereby saving the plant operator those costs associated with downtime of the reactor. Should the openings 19 within the cooling shroud 16 at any particular plant be viewing windows instead of access doors 29, the monitoring system 25 can be secured thereto in an alternative manner. The video cameras 28 and lamps 46, 47 may be secured to the exterior surface of an adapter plate (not shown) which replaces the existing viewing windows. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alterations would be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention, which is to be given by the full breadth of the appended claims and in any and all equivalents thereof. |
claims | 1. A neutron capture therapy system, comprising a beam shaping assembly, wherein the beam shaping assembly comprises,a beam inlet;a neutron generator arranged in the beam shaping assembly, wherein the neutron generator has nuclear reaction with an incident proton beam from the beam inlet to produce neutrons;a moderator adjoining to the neutron generator, wherein the neutrons are moderated by the moderator to epithermal neutron energies;a reflector surrounding the neutron generator and the moderator, wherein the reflector leads the deflected neutrons back to enhance the epithermal neutron beam intensity;a beam outlet; andat least a movable member moving away from or close to the neutron generator, wherein the movable member moves between a first position where the neutron generator is replaceable and a second position where the neutron generator is irreplaceable. 2. The neutron capture therapy system according to claim 1, wherein the movable member is part of the reflector or a combination of part of the reflector and part of the moderator. 3. The neutron capture therapy system according to claim 1, wherein the reflector defines a first linear height, the movable member defines a second linear height, the neutron generator defines a third linear height, and the second linear height is smaller than or equal to the first linear height and is greater than the third linear height. 4. The neutron capture therapy system according to claim 3, wherein the movable member defines a symmetry plane, the movable member is symmetrical with respect to the symmetry plane, and the symmetry plane passes through an axis of the neutron generator. 5. The neutron capture therapy system according to claim 2, wherein the reflector includes a first connecting part which comprises a first connecting surface and a first junction surface; the movable member includes a second connecting part which comprises a second connecting surface and a second junction surface; the first connecting surface connects with the second connecting surface and is overlapped with the second connecting surface, the first junction surface connects with the second junction surface and is overlapped with the second junction surface, and the second junction surface is overlapped with the plane where the axis of the neutron generator is in. 6. The neutron capture therapy system according to claim 3, wherein when the second liner height is smaller than the first linear height and greater than the third height, the first connecting surface and the first junction surface are overlapped, the second connecting surface and the second junction surface are overlapped; when the second linear height is equal to the first linear height, the first connecting surface and the first junction surface are connected but are located in different planes, the second connecting surface and the second junction surface are connected but are located in different planes, the first connecting surface connects with the second connecting surface and is overlapped with the second connecting surface, and the first junction surface connects with the second junction surface and is overlapped with the second junction surface. 7. The neutron capture therapy system according to claim 1, wherein the neutron capture therapy system further comprises a driving assembly; the driving assembly comprises at least a gate to support the movable member, and a guide rail for allowing the gate to move away from or close to the neutron generator; when the gate moves away from the beam shaping assembly, the movable member moves to the first position; and when the gate moves close to the beam shaping assembly, the movable member moves to the second position. 8. The neutron capture therapy system according to claim 7, wherein the guide rail is arranged outside the beam shaping assembly, the gate moves away from or close to the beam shaping assembly along the guide rail. 9. The neutron capture therapy system according to claim 6, wherein the guide rail is arranged outside the beam shaping body, the driving assembly further comprises supporting a frame, one end of the supporting frame is used for supporting the gate, the other end of the supporting frame moves in the guide rail, and the gate moves away from or close to the beam shaping assembly as the supporting frame moves along the guide rail. 10. The neutron capture therapy system according to claim 2, wherein when the second linear height is smaller than the first linear height and is greater than the third linear height, a guide rail is arranged in the reflector, and the movable member is arranged in the beam shaping assembly and moves away from or close to the neutron generator through the guide rail. 11. A neutron capture therapy system, comprising a beam shaping assembly, wherein the beam shaping assembly comprises,a beam inlet,a neutron generator arranged in the beam shaping assembly, wherein the neutron generator has nuclear reaction with an incident proton beam from the beam inlet to produce neutrons;a moderator adjoining to the neutron generator, wherein the neutrons are moderated by the moderator to epithermal neutron energies;a reflector surrounding the moderator, wherein the reflector leads the deflected neutrons back to enhance the epithermal neutron beam intensity;a beam outlet; andat least a movable member moving away from or close to the neutron generator, wherein the movable member moves between a first position where the neutron generator is replaceable and a second position where the neutron generator is irreplaceable, and wherein the movable member defines a curved surface, when the moveable member moves away from the beam shaping assembly, a notch is shaped on the beam shaping assembly where the neutron generator is replaced; and when the moveable member moves close to the beam shaping assembly, the curved surface surrounds at least half of the neutron generator. 12. The neutron capture therapy system according to claim 11, wherein the movable member is a part of the reflector or a combination of part of the reflector and part of the moderator. 13. The neutron capture therapy system according to claim 11, wherein the reflector defines a first linear height, the movable member defines a second linear height, the neutron generator defines a third linear height, and the second linear height is smaller than or equal to the first linear height and is greater than the third linear height. 14. The neutron capture therapy system according to claim 11, wherein the movable member defines a symmetry plane, the movable member is symmetrical with respect to the symmetry plane; the symmetry plane passes through an axis of the neutron generator. 15. The neutron capture therapy system according to claim 11, wherein the neutron capture therapy system further comprises a driving assembly; the driving assembly comprises at least a gate for supporting the movable member, and a guide rail for allowing the gate to move away from or close to the neutron generator; the guide rail is arranged outside the beam shaping assembly, and the gate moves away from or close to the beam shaping assembly through the guide rail, the movable member moves away from or close to the neutron generator along the guide rail. 16. The neutron capture therapy system according to claim 15, wherein the driving assembly further comprises a supporting frame, one end of the supporting frame supports the gate, the other end of the supporting frame moves in the guide rail, and the gate moves away from or close to the beam shaping assembly as the supporting frame moves along the guide rail. 17. The neutron capture therapy system according to claim 11, wherein when the second linear height is smaller than the first linear height and is greater than the third linear height, a guide rail is arranged in the reflector, and the movable member is arranged in the beam shaping assembly and moves away from or close to the neutron generator through the guide rail. 18. A neutron capture therapy system, comprising a beam shaping assembly, wherein the beam shaping assembly comprises,a beam inlet;a neutron generator arranged in the beam shaping assembly, wherein the neutron generator has nuclear reaction with an incident proton beam from the beam inlet to produce neutrons;a moderator adjoining to the neutron generator, wherein the neutrons are moderated by the moderator to epithermal neutron energies;a reflector surrounding the moderator, wherein the reflector leads the deflected neutrons back to enhance the epithermal neutron beam intensity, and wherein the reflector includes a first connecting surface and a first junction surfaces connects to the first connecting surface;a beam outlet; andat least a movable member moving away from or close to the neutron generator, wherein the movable member moves between a first position where the neutron generator is replaceable and a second position where the neutron generator is irreplaceable, and wherein the movable member defines a second connecting surface and a second junction surface connected with the second connecting surface, the second connecting surface is overlapped with the first connecting surface, the second junction surface is overlapped with the first junction surface, when the moveable member moves away from the neutron generator, the first connecting surface is separated from the second connecting surface, the first junction surface is separated from the second junction surface, and the neutron generator is replaceable. 19. The neutron capture therapy system according to claim 18, wherein the first connecting surface is inclined to the first junction surface, the second connecting surface is inclined to the second junction surface. 20. The neutron capture therapy system according to claim 18, wherein the second junction surface includes a recess, the first junction surface includes a protrusion for engaging to the recess in a form-fitting manner. |
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abstract | A method of reducing the volume of a blade section of a boiling water reactor control rod for transport or storage that cuts the control rod spline into four substantially equal longitudinal sections, with each longitudinal section including one control rod blade. Each longitudinal section is radiologically characterized and the locations of desired lateral segmentation are identified. A band of malleable metal is wrapped around each longitudinal section at each of the locations and the bands are sheared to separate segments of the longitudinal section and the ends of the bands are crimped at the point of shearing to seal the interior of the segments. |
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claims | 1. A method of making an anti-scatter X-ray grid device comprising:providing a substrate comprising a first material substantially non-absorbent of X-rays, the substrate having a plurality of channels therein;applying a layer onto a sidewall of the plurality of channels, wherein the layer comprises a second material substantially non-absorbent of X-rays; andapplying a third material substantially absorbent of X-rays into a portion of the plurality of channels, thereby defining a plurality of X-ray absorbing elements. 2. The method of claim 1 wherein the second material is a conformal coating. 3. The method of claim 1 wherein the third material comprises at least one of lead, tungsten, uranium, gold, a polymer containing lead, a polymer containing tungsten, and a polymer containing gold. 4. The method of claim 1, wherein a length or width of the anti-scatter X-ray grid device is in a range from about 12 cm to about 40 cm. 5. The method of claim 1, wherein a width of the plurality of X-ray absorbing elements is less than about 20 μm. 6. The method of claim 1, wherein a width of the plurality of X-ray absorbing elements is in a range from about 5 μm to about 10 μm. 7. The method of claim 1, wherein the applying a layer comprises applying a conformal coating in the plurality of channels. 8. The method of claim 7, wherein the conformal coating comprises an oxide, a nitride, a plastic, a polymer, an acrylic, an epoxy, a urethane, silicone, and combinations thereof. 9. The method of claim 7, wherein the conformal coating comprises Parylene. 10. The method of claim 1, wherein the plurality of X-ray absorbing elements are configured in an angular orientation. 11. The method of claim 1, further comprising prior to providing a substrate, making the plurality of channels in the substrate by at least one of injection molding, laser, mechanical removal, and plasma etching. 12. The method of claim 1, wherein the applying a layer comprises applying a layer on both sidewalls of the plurality of channels. 13. The method of claim 12, wherein the applying a layer comprises applying a conformal coating to a surface of the plurality of channels. 14. The method of claim 13, wherein the conformal coating is applied by one of vacuum deposition, evaporation, chemical vapor deposition, and sputtering. 15. The method of claim 1, wherein the applying a third material comprises filling the plurality of channels. 16. The method of claim 1, further comprising planarizing a top surface of the grid device. 17. An anti-scatter X-ray grid device comprising:a substrate comprising a first material substantially non-absorbent of X-rays, the substrate having a plurality of channels therein;a second material substantially non-absorbent of X-rays, lining sidewalls of the plurality of channels; anda third material substantially absorbent of X-rays, at least partially resident in the plurality of channels, thereby defining a plurality of X-ray absorbing elements. 18. The anti-scatter X-ray grid device of claim 17, wherein a width of the plurality of X-ray absorbing elements is less than about 20 μm. |
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051777746 | summary | FIELD OF THE INVENTION The field of the present invention relates generally to microscopy, and more particularly to X-ray microscopy. BACKGROUND OF THE INVENTION The use of X-ray radiation in microscopy is known in the prior art. Such known X-ray microscopes typically are of the transmission mode type microscope, which include the source of X-rays on one side of the specimen or object to be viewed, and a radiation detector or imaging system on the other side of the object. Accordingly, only X-rays going through the material are detected. As a result, the use of such transmission type X-ray microscopes for viewing non-transparent or poorly transparent objects is limited, particularly with respect to soft X-ray lasers. The technology related to X-ray microscopy has been greatly improved over the last ten years. Significant progress has been made in developing X-ray lasers, and synchrotron insertion devices, for providing increased brightness of laboratory X-ray sources. Optical microscopy is limited to a resolution of about 3,000 angstroms, or about 0.3 micrometer. In lithographic systems used for inspecting integrated circuits, the capability for inspecting line widths of such circuits of less than 0.1 micrometers and below is now required. Electron microscopes can provide such resolution, but electron microscopes are slow to use for such inspection, and the electron beam can cause functional damage to the integrated circuit chip. Skinner et al., in a paper entitled "Contact Microscopy With A Soft X-ray Laser", appearing in the Journal of Microscopy, Volume 159, part I, July, 1990, on pages 51 through 60, discusses the use of soft X-ray lasers for imaging live cells at high resolution, "thereby bridging the gap between electron microscope images of non-live cells that have undergone extensive specimen preparation, and low resolution but high fidelity images of live cells recorded with light microscopes." It is recognized that "to be of maximum utility to biologists, a soft X-ray laser contact microscope should be suitable for everyday use on fragile, living biological specimens." The paper also describes a system developed at Princeton University, Princeton, N.J., for generating soft X-ray laser beams. The installation of a contact microscope on a soft X-ray laser beam is shown. In this system, the soft X-ray laser beam is used to cause the shadow of a specimen to be recorded on photo-resist material. Howells, et al., in a paper entitled "X-ray Microscopes", Scientific American, Volume 264, No. 2, February, 1991, pages 88 through 94, describes the state of X-ray microscopy, and the ongoing development of "soft" X-ray instruments, providing more than ten times better resolution than optical microscopes. It is indicated that "soft X-rays in the wavelength range of 20.0 to 40.0 angstroms (an angstrom is one ten-billionth of a meter), are sufficiently penetrating to image biological cells in many cases." The article goes on to describe the difficulty of focusing X-ray images and developments which led to the construction of a Fresnel zone plate for providing such focusing. Contact microradiography is described, as previously mentioned above, whereby an X-ray beam is passed through a sample to a resist (PMMA) material, causing a damage pattern in the resist material relating to details of the sample. Imaging X-ray microscopes are described whereby X-ray beams are passed through a condenser zone plate, for focusing the X-rays onto a region of a sample. The X-rays pass through the sample onto a micro-zone plate for focusing the X-rays into an image field that is picked up by a detector. A scanning X-ray microscope system is described whereby an X-ray beam is focused and scanned back and forth across a sample, with the scan beam being passed through the sample and detected by an X-ray counter. A soft X-ray laser contact microscope system is described in Suckewer et al. U.S. Pat. No. 4,979,203. The described system uses an optical contrast microscope for inspecting and aligning a target prior to applying soft X-rays to the target for performing contact X-ray microscopy. An X-ray laser is used for providing the necessary soft X-ray beam. Fields, et al. U.S. Pat. No. 3,702,933, entitled "Device and Method for Determining X-ray Reflection Efficiency of Optical Surfaces", teaches a method for determining the X-ray reflection efficiency and scattering characteristics of optical surfaces using much harder (shorter wavelength) X-ray radiation than the soft X-rays described above. As shown in the figure, X-rays of a known wavelength are generated by an X-ray source 11, and passed through a collimator 15, and therefrom passed through slits 17, for projection onto an area of a crystal monochromator 21 for diffracting the X-rays. The diffracted x-rays are then passed through slits 25, and projected onto a predetermined area of an optical test specimen 41. X-rays reflected off of the optical test specimen 27, are transmitted through a slit 33, and projected therefrom into an X-ray detector 31. The X-ray detector can be a Geiger-Muller counter. The intensity of the X-rays prior to and subsequent to reflection from the specimen 27 are compared for determining the efficiency of reflection of the optical surface of the test specimen 27. Suckewer et al. U.S. Pat. No. 4,771,430 shows an apparatus for enhancing soft X-ray lasing action through use of thin blade radiators in a target. Soft X-ray lasing action is generated in a defined plasma column. The plasma is produced by focusing a CO.sub.2 laser beam onto a carbon target. A magnetic field is used to compress the plasma into a thin column. A carbon disc in combination with carbon blades mounted perpendicular to the surface of the disc provides the target. The CO.sub.2 laser beam is directed to strike the surface of the disc for forming a plasma column. The column is cooled by radiation losses and heat conduction to the blade. The resulting soft X-rays are transmitted through a slot in the carbon disc. Rocca U.S. Pat. No. 4,937,832 shows a method and apparatus for producing a soft X-ray laser beam in a capillary discharge plasma. As shown in FIGS. 1 and 2 thereof, the apparatus includes a pair of electrodes having axially oriented poles, respectively, for facilitating the exit of laser radiation. The electrodes are connected to a discharge circuit that includes a relatively large capacitor that is first charged to a given level, and then discharged across the electrodes. It is indicated that the power source can also be an electrical transmission line having a low impedance. Also discussed is the use of a high intensity magnetic field for containing the plasma. Another Suckewer U.S. Pat. No. 4,704,718, similar to the above-described Suckewer patent, teaches the creation of a plasma column by focusing a CO.sub.2 laser pulse on a carbon target. A magnetic field is used to contain the plasma. There are many other examples of soft X-ray generators and/or ultraviolet X-ray generators in the art. Other such references include U.S. Pat. Nos. 4,555,787, and 3,956,711. Research is ongoing for providing improved sources of optics at short wavelengths, particularly wavelengths in the soft X-ray region from 1.0 nanometer to 30.0 nanometers wavelength. As previously indicated, laser sources and synchrotrons have provided soft X-rays for use in transmission X-ray microscopy. X-ray lasers provide a very high flux of short wavelength photons in very short pulses in single lines, whereas synchrotrons provide continuously tunable radiation. These sources of X-ray radiation tend to complement one another. The present inventors recognized that by developing a reflection X-ray laser microscope, a step forward can be made in the field of the art for using such an apparatus in the field of lithography for inspecting integrated circuits, for example, and/or in the medical fields for analyzing biological specimens. They further recognized that the reflection coefficient for a number of biological materials is significant and differs substantially at soft X-ray wavelengths. They recognized that with a high flux of radiation from a soft X-ray laser, and through use of high sensitivity CCD (charge coupled device) detectors, a very compact reflection soft X-ray microscope can be provided. The inventors expect that such a microscope will provide magnification up to .times.100 with the resulting images recorded on CCD-array detectors, for example, having a pixel size in the order of 5.0 microns. Such a soft X-ray reflecting microscope is expected to have a resolution in the range of 0.05 microns. It is further expected a soft X-ray laser source for use with the reflection X-ray microscope of the present invention will operate at a wavelengths of 18.2 nanometers (nm) with a beam energy of 1.0 to 3.0 mJ (millijoule) in a 10.0 to 30.0 nanosecond pulse. The laser can operate at shorter wavelengths (15.4 nm and 12.9 nm) with lower beam energy. A pumping CO.sub.2 laser pulse with an energy of 300.0 to 500.0 joules in 50.0 to 75.0 nanoseconds will create a lasing medium (plasma column) in a strong, magnetic field. Also, the repetition rate of the X-ray laser is three minutes. The present inventors are scientists at the Princeton University Plasma Physics Laboratory (PPPL) (Skinner and Suckewer) and Mechanical and Aerospace Engineering Department (Suckewer) and Princeton X-ray Laser Inc. (Rosser), in Princeton, N.J., where soft X-ray laser development has been pursued for a number of years. In the March, 1987, issue of the "PPPL Digest", published by the Information and Administrative Services, Princeton Plasma Physics Laboratory, Princeton, N.J., the basics of soft X-ray technology existing in 1987 and various experimental results are discussed. In a later May, 1989 issue of the "PPPL Digest", a description of X-ray laser microscopy research at that facility is described. A composite X-ray laser microscope is shown and described for soft X-ray laser contact microscopy. Also, the basic principals of a soft X-ray laser are shown and described. SUMMARY OF THE INVENTION An object of the invention is to provide a reflecting soft X-ray microscope. With the problems of the prior art in mind, these and other objects are provided in one embodiment of the invention by positioning a source of soft X-rays (for example, a soft X-ray laser) for directing a soft X-ray beam through focusing means for focusing the beam to strike an object or specimen at a predetermined angle, for illuminating a portion of the specimen, a portion of the beam reflected off of the object o specimen is passed into imaging means for focusing the reflected beam, for forming an image for detection by a CCD (charge coupled device) array or X-ray sensitive film, for example. |
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description | The present invention relates to the field of the treatment of nuclear waste. It relates in particular to the treatment of waste containing sodium and at least one radioactive substance. One such waste is for example a pin for controlling the reactivity of a sodium-cooled Fast Neutron Reactor (“FNR-Na”). In order to control nuclear reactivity, “FNR-Na” reactors use a neutron absorbing material containing boron carbide of simplified formula B4C. This material is generally in the form of sintered cylindrical pellets stacked in a cladding, in order to form an absorbing element such as an absorber pin. Under the combined action of temperature and irradiation, the initially massive boron carbide pellets may be degraded until cracks appear in the pellets. During operation of the “FNR-Na” reactor, the sodium in the primary circuit is in liquid form and contains at least one radioactive substance. It circulates in the space between the boron carbide pellets and the cladding. Following degradation of the pellets, the liquid sodium contaminated with the radioactive substance may then penetrate into the cracks in the boron carbide pellets, or even along the fractures of the boron carbide pellets when the cracks have led to fragmented pellets. In the present description, the fractures are classed as cracks. After stopping the reactor, the absorber pins are extracted from the reactor and then put in storage before treatment. The absorber pin then comprises cracked boron carbide pellets, in which the cracks contain sodium which is in solid form and is contaminated with at least one radioactive substance. The contaminated absorber pin constitutes nuclear waste that poses a dual risk in terms of safety and security: a chemical risk due to the residual sodium, which must be kept under an inert gas (such as argon or nitrogen) so that there is no risk of chemical reaction, for example with water or with the oxygen of the air. Depending on the conditions of storage before treatment, a proportion of the sodium at the surface may nevertheless be transformed, for example to soda and to hydrogen on contact with water, and in an uncontrolled manner; a radiological risk due to contamination of the sodium with the radioactive substance, namely the radioactive isotopes from the primary circuit of the reactor. In order to be able to treat such nuclear waste through the conventional channels for removal of contaminated waste, it is firstly necessary to eliminate the chemical risk, i.e. to transform chemically or extract the contaminated metallic sodium present in the absorber pin, in particular in the cracks in the boron carbide pellets. A method for chemical transformation of sodium by direct reaction between water and sodium is difficult to implement: it requires bringing these two chemical species into contact, but also requires control of the reaction kinetics, disposal of the soda and hydrogen produced, as well as absence of accumulation of reagents. Now, sintered boron carbide is a chemically stable material of low porosity, the porosity generally representing less than 1% of the volume of the material. The sodium is therefore very confined within this material. Consequently, numerous cutting operations on the cladding would be required for treating the sodium-boron carbide radioactive mixture. However, this operation is long and difficult, as the hardness of the boron carbide pellets is such that it would damage and contaminate the cutting tools. The presence of a radioactive substance also means working in a confinement enclosure under inert gas, such as a glove box. Now, the cutting operations are also difficult there, because of the difficulties of manipulation intrinsic to this type of enclosure. Moreover, they would generate dispersion of radioactive substance in the enclosure, which must be limited as far as possible. The poor accessibility of the contaminated sodium therefore complicates its treatment as waste considerably. One of the aims of the invention is therefore to avoid or attenuate one or more of the drawbacks described above, by carrying out a method which among other things allows easy treatment of the sodium and radioactive substance contained in the cracks of a material based on sintered boron carbide, which is a chemically stable material of great hardness, the sodium to be treated being difficultly accessible because it is contained within the cracks. The invention thus relates to a method for treating an absorber pin, the pin comprising a cladding in which there is a material based on sintered boron carbide that has cracks, the cracks containing sodium and at least one radioactive substance. The method comprises a treatment step, during which the sodium is converted to sodium carbonate by a carbonation reaction by contacting the material with a treatment reaction mixture comprising in molar percentage 0.5% to 5% of steam, 5% to 25% of carbon dioxide and 74.5% to 94.5% of a chemically inert gas, in such a way that expansion of the carbonate causes opening up of the cracks and of the cladding starting from at least one slit made in the cladding as well as propagation of the method of treatment within the material. The treatment step comprises contacting the material with a treatment reaction mixture in order to carry out a carbonation reaction in which the soda obtained by a hydrolysis reaction, after contacting steam with sodium, is converted to carbonate after reaction with the carbon dioxide contained in the treatment reaction mixture. The treatment reaction mixture comprises reagents in gaseous form. It therefore comes into contact more easily with the difficultly accessible sodium that is contained in the cracks in the boron carbide pellets, or even with the sodium present on the fractures or on the interfaces of the individual pellets of sintered boron carbide. This performance cannot be achieved by a method of treatment using only water, even in large amounts. The carbonation reaction employed according to the treatment step generates a carbonate composed essentially of sodium carbonate Na2CO3 and/or sodium hydrogen carbonate NaHCO3, which in the present description is classed with the sodium carbonate. The volume occupied by the carbonate is greater than the volume initially occupied by the sodium. The carbonation reaction is then carried out in such a way that the volume expansion of the carbonate advantageously causes widening of the cracks and opening up of the cladding, the latter having previously been weakened mechanically by making at least one slit. The openings thus produced open the way for continuation of the carbonation reaction and its propagation to zones that were initially inaccessible to the reagents, namely water and carbon dioxide. All of the sodium contained in the cladding can then be treated. The slit is preferably longitudinal and/or made on the entire length of the cladding. As the cladding is generally made of metal, most often composed of stainless steel, the slit is for example made using a laser. If applicable, the ends of the cladding may be cut along a transverse plane in order to increase the kinetics of the chemical reactions employed during the steps of pretreatment and treatment. Following opening up of the cracks and of the cladding, the carbonation reaction according to the treatment step continues by gradual increase in the area of contact between the sodium and the gaseous reaction mixture. Thus, opening up of the cracks and of the cladding causes acceleration of the carbonation reaction, which can then be propagated into the whole of the cladding so as to treat all of the sodium. This result is obtained despite the fact that the cracks are not generally connected and constitute confined reaction spaces. These spaces prevent or limit a priori the propagation of the carbonation reaction, or of the hydrolysis reaction using only water, which is envisaged conventionally for this type of treatment. Despite the poor initial accessibility of the sodium contained in the cracks in the material, the treatment reaction mixture can react in depth and completely with the contaminated sodium. The cladding can then be treated by means of the method of the invention without the need for numerous cutting operations. This is particularly advantageous, because owing to the presence of a radioactive substance, the method of treatment according to the invention is most often carried out in a confinement enclosure such as a glove box, a hot cell or a chemical reactor, in which, as noted above, one tries to limit the cutting operations. Moreover, through the use of the carbonation reaction, the method of treatment according to the invention has the advantage that it only produces solid waste (sodium carbonate, boron carbide, cladding and radioactive substance) and gaseous waste (hydrogen). Therefore no liquid or gaseous radioactive effluent is generated. Sodium carbonate is a product that is stable and inert. It is easy to handle and is compatible with the final outlet channels for the boron carbide pins. It can be stored directly in a long-term deep repository. The hydrogen produced in the carbonation reaction may be removed by a scavenging gas. The method of treatment according to the invention is also easily controllable since the carbonation reaction according to the treatment step can be slowed down by decreasing the proportion of steam in the gaseous treatment reaction mixture, or even stopped by replacing this mixture with inert gas. This makes the method of the invention extremely safe. Associated with its relative simplicity of use, the method of the invention also makes it possible to treat a larger number of absorber pins in a single operation, which constitutes an important economic advantage. An additional difficulty may nevertheless arise when a crust comprising soda (NaOH) and/or sodium oxide (Na2O) covers at least a portion of the surface delimiting the cracks in the material. Such a crust may form by hydrolysis of the sodium in the presence of water, and then concentration and crystallization of the products obtained. It then covers the sodium present in the cracks and constitutes a protective layer preventing contacting of the treatment reaction mixture with the underlying sodium, and therefore propagation of the carbonation reaction in order to treat the contaminated sodium present under the crust. For this reason, according to a preferred embodiment of the method of treatment according to the invention, a pretreatment step is carried out in which the material is contacted with a pretreatment reaction mixture comprising in molar percentage 0.5% to 25% of carbon dioxide and a chemically inert gas for the remainder. The pretreatment step precedes the treatment step. The pretreatment reaction mixture is a dry mixture that does not contain water or contains small amounts of water, so that the carbon dioxide that it contains destroys the crust by converting the soda and/or the sodium oxide to sodium carbonate. The small amounts of water permissible are therefore those that prevent growth of the crust that is greater than its destruction in the pretreatment step. It comprises a chemically inert gas that has the same properties as the treatment reaction mixture. Any gas that is chemically inert with respect to sodium may be suitable. The inert gas is for example nitrogen, argon or a mixture thereof. The pretreatment reaction mixture and the treatment reaction mixture may be identical. The chemical composition of the treatment reaction mixture therefore differs essentially from that of the pretreatment reaction mixture by the additional presence of water. This results in simplification of the installation in which the method of treatment according to the invention is carried out, such as a confinement enclosure for example. The number of lines for introducing the reagents into the enclosure is thus limited. This reinforces the confinement of the enclosure and therefore the robustness and safety of the method of treatment despite the presence of a radioactive substance. The pretreatment step produces sodium carbonate, which is a compound that is also obtained at the end of the treatment step. The chemical composition of the waste obtained at the end of the treatment and pretreatment steps is thus limited. This advantageously reduces the number of disposal channels and of operations necessary for the subsequent treatment of the waste. In the present description of the invention, verbs such as “comprise”, “contain”, “incorporate”, “include” and their conjugated forms are open terms and therefore do not exclude the presence of additional element(s) and/or step(s) that may be added to the initial element(s) and/or step(s) stated after these terms. However, these open terms further relate to a particular embodiment in which only the initial element(s) and/or step(s), excluding all others, are aimed at; in which case the open term further entails the closed term “consist of” or “compose of” and its conjugated forms. The expression “and/or” is understood to link elements with a view to denoting their individual presence, but also a mixture or combination thereof. Moreover, unless stated otherwise, the values at the limits are included in the ranges of parameters indicated. Despite the presence of a radioactive substance, for example cesium or tritium (fission products), cobalt 60 or manganese 54 (activation products), the method of the invention treats the sodium present in the cracks in the material based on sintered boron carbide. This material is generally in the form of pellets. It is constituted wholly or partly of sintered boron carbide, the composition of which in carbon atoms is generally between 8.8% and 20%, and may therefore optionally vary in this range relative to the stoichiometric formula B4C corresponding to 20% of carbon atoms, or even have an excess of carbon up to 1 wt %. The method of treatment according to the invention comprises a treatment step during which the material is brought into contact with a treatment reaction mixture comprising in molar percentage 0.5% to 5% of steam, 5% to 25% of carbon dioxide and 75% to 94.5% of a chemically inert gas. The minor proportion of water in the treatment reaction mixture prevents any condensation of water on the walls of the cladding and thus allows treatment of the sodium in complete safety. The time of contact of the treatment or pretreatment reaction mixture with the material depends respectively on the amount of crust or sodium to be treated, or also on the composition of the reaction mixture. A person skilled in the art can easily adapt this time in order to obtain treatment of the sodium contained in the absorber pin that is as complete as possible, which is indicated for example by the end of release of hydrogen. Contacting with the treatment or pretreatment reaction mixture is for example carried out for a duration between 5 hours and 15 days. It is preferably carried out at a temperature between 40° C. and 55° C. Regarding the treatment step, this prevents the water condensing, even at the maximum concentrations of 5 mol % of steam, and reacting violently with the sodium. Since the material contains at least one radioactive substance, the treatment according to the invention is most often carried out in a confinement enclosure such as a glove box or a hot cell. The treatment or pretreatment reaction mixture is then generally introduced into the confinement enclosure at a flow rate allowing its continuous renewal at least once per hour, typically once to twice per hour. Other aims, features and advantages of the invention will now be stated in the following description of a particular embodiment of the method of the invention, given for purposes of illustration, and not limiting. The ends of a metal cladding comprising sintered boron carbide pellets that have cracks and that contain sodium and a radioactive substance are cut off using a laser. Then a longitudinal slit is made in the metal cladding. In a glove box thermostatically controlled at 45° C., the pellets are brought into contact with a pretreatment reaction mixture with the aim of removing the crust of soda that forms on the surface of the cracks. This mixture contains, in molar percentage, 10% of carbon dioxide and 90% of nitrogen. The pellets are then brought into contact with a treatment reaction mixture containing, in molar percentage, 3% of steam, 10% of carbon dioxide and 87% of nitrogen. After some hours, absence of release of hydrogen indicates the end of treatment. The solid waste obtained, namely sodium carbonate, boron carbide, the cladding and the radioactive substance, may be packaged in order to be removed to the appropriate channels. |
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abstract | An ion transport apparatus for a mass- or ion-mobility-spectrometer comprises: (a) a plurality of strip electrodes in a series on a flat substrate; (b) an ion outlet aperture in the substrate disposed adjacent to a first one of the plurality of strip electrodes; (c) a cage electrode at least partially enclosing the plurality of strip electrodes and the ion outlet aperture; (d) a radio frequency (RF) voltage generator operable to supply an RF phase difference between each pair of adjacent electrodes; and (e) at least one DC voltage source operable to supply first and second DC voltages to the cage electrode and an extraction electrode and to supply respective DC bias voltages to each of the plurality of electrodes, wherein electrode strip widths of a series of the plurality of electrodes progressively increase away from the first one of the plurality of electrodes. |
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039740277 | claims | 1. A water-cooled nuclear reactor installation comprising a metal pressure vessel having a vertical substantially cylindrical side wall having an upper portion and coolant pipes radiating from said upper portion, a concrete pit in which said vessel is positioned and forming a concrete wall surrounding said vessel at a radial distance therefrom defining an annular space around the vessel, said space when empty having a radial extent permitting instrumentation to be lowered into said space for inspecting the vessel's said side wall, said space being substantially filled by a rupture-protecting encasement for the vessel's said wall; said encasement comprising a shell formed by individual-separable segmentally-cylindrical segments interfitted to form at least one substantially cylindrical layer encasing the vessel's said side wall, said segments being made of pressure-resistant heat-insulating non-metallic material and the segments subdividing said layer in the radial and axial direction of the vessel's said side wall and said shell being encircled by high-tensile strength metal rings; said metal pressure vessel radially expanding thermally when said vessel is hot and said shell thermally insulating said metal rings against heat from said vessel so that said rings do not correspondingly thermally expand, said rings therefore applying radial compression via said shell to the vessel's said side wall to provide said rupture-protecting encasement when said vessel is hot, said metal pressure vessel thermally contracting radially when cool and said metal rings then releasing said compression, whereby to permit upward removal of said segments, segment-by-segment, when said vessel is cool, and freeing said space from said segments, to permit said inspection. 2. The installation of claim 1 in which said shell comprises a plurality of said layers which are radially superimposed relative to each other between said rings and the pressure vessel's said wall. 3. The installation of claim 1 in which said pit and wall have radial holes through which said pipes radiate, said holes having walls forming radial spaces around said pipes, the last-said spaces being substantially filled by rupture-protecting encasements for said connections, each encasement comprising interfitted individually-separable segmentally-cylindrical segments of said non-metallic material forming layers encircling said connections, and high-tensile-strength metal rings encircling the just-named layers and correspondingly applying via the just-named layers radial compression to said pipes when hot. |
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059303158 | abstract | A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced. |
claims | 1. A method for removing radionuclides adhered to the surface of at least one of stainless steel or aluminum material, the method comprising the steps of:a) contacting the material with a carbonate/bicarbonate electrolyte solution having a pH of about 4, the electrolyte solution containing at least one of sodium or potassium ions; andb) electrolytically removing the radionuclides from the surface of the metal whereby radionuclides are caused to be stripped off of the material without corrosion or etching to the material surface. 2. The method for removing radionuclides, as defined in claim 1, wherein the carbonate concentration is about 2% by weight of the bicarbonate. 3. The method for removing radionuclides, as defined in claim 1, wherein the radionuclides are selected from the group consisting of cesium, strontium and actinides. 4. A method for reclaiming radiation contaminated equipment constructed from at least one of stainless steel or aluminum material, the method comprising the steps of:a) providing an electrolytic treatment vessel, the treatment vessel including a cathode and an electric current power supply for supplying DC current thereto;b) providing an electrolytic solution within the treatment vessel, the electrolytic solution comprising a carbonate/bicarbonate solution having a pH of about 4 and containing at least one of sodium or potassium ions;c) positioning equipment to be reclaimed in the tank and connecting the equipment to the power supply;d) selectively applying a DC current between the cathode and the equipment for a period of time sufficient to cause electrolytic removal of radionuclides from the surface of the equipment whereby radionuclides are stripped off of the equipment without corrosion or etching to the surface of the equipment;e) washing the equipment following electrolysis; andf) recovering the electrolytic solution for further treatment. 5. The method for reclaiming radiation contaminated equipment, as defined in claim 4, further including the step of distilling the recovered solution to separate the water from the stripped radionuclides. 6. The method for reclaiming radiation contaminated equipment, as defined in claim 4, wherein the carbonate concentration is about 2% by weight of the bicarbonate. 7. The method for reclaiming radiation contaminated a equipment, as defined in claim 4, wherein the radionuclides are selected from the group consisting of cesium, strontium and actinides. |
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abstract | Disclosed is a nuclear fuel rod including at least one or more fuel pellets, a cladding tube surrounding the fuel pellets, and burnable absorber inside the cladding tube. The burnable absorber comprises a burnable absorber material and a cladding material surrounding the burnable absorber material. The burnable absorber has a disk shape, and the cladding material is an alloy comprising zirconium. |
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abstract | A method for analyzing a sample by diffractometry and a diffractometer, where the diffractometer includes a collimated source, a detection collimator, and a spectrometric detector, the detection axis of the detector and the collimator form a diffraction angle with the central axis of an incident beam and an energy spectrum is established for each pixel of the detector. The measured spectra are readjusted by a change in variable that takes into account the energy of the scattered radiation and the angle of observation. The measured are combined and a check is made on the implementation of at least one multi-material criterion representative of the presence of a plurality of layers of materials and groups of pixels are formed according to the results of this check, where each group corresponds to a single layer of material and the measured spectra obtained for the pixels of the group are combined. |
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050664514 | summary | CROSS REFERENCE TO RELATED APPLICATION Reference is hereby made to the following copending patent application dealing with related subject matter and assigned to the assignee of the present invention: "Curvilinear Translating Latch And Linkage Arrangement In A Control Rod Drive Mechanism Of A Nuclear Reactor" by John E. Tessaro, assigned U.S. Ser. No. 567,026 and filed Aug. 14, 1990. BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactors and, more particularly, is concerned with a method of operating a control rod drive mechanism for performing single-step multiple repositionings of a control rod cluster assembly during a single fuel cycle of the nuclear reactor. 2. Description of the Prior Art In a commercial nuclear reactor, heat, from which steam and ultimately electricity are generated, is produced by fissioning of a fissile material such as enriched uranium. This fissile material, or nuclear fuel, is typically contained within a nuclear core made up of a multiplicity of fuel rods supported in a plurality of nuclear fuel assemblies, coextensively arranged in a spaced parallel array. Movable control rods are dispersed throughout the core to control the fission process. The control rods generally comprise a plurality of elongated rods containing neutron absorbing materials which fit in longitudinal openings defined in the fuel assemblies and among the fuel rods by guide thimbles of the fuel assemblies. The guide thimbles thus guide the control rods during their movement into and out of the core. Inserting a control rod into the core adds more absorber material and, hence, decreases the nuclear reaction; conversely, withdrawing a control rod removes absorber material and, hence, increases a nuclear reaction and thereby the power output of the core. The nuclear reactor core and the control rods are positioned within and supported by a reactor vessel through which a reactor coolant flows. The control rods are supported in cluster assemblies moved into and from the nuclear core by control rod drive mechanisms which, in turn, are mounted by an upper internals arrangement located within the nuclear reactor vessel above the nuclear core. Typically, a reactor pressure vessel is pressurized to a relatively high internal pressure. The control rod drive mechanisms operate within the same pressure environment that exists within the reactor pressure vessel. Hence, the control rod drive mechanisms are housed within pressure housings of the upper internals arrangement which are tubular extensions of the reactor pressure vessel. One of the more commonly used types of control rod drive mechanisms is referred to as a "magnetic jack" With this type of mechanism, the control rods are jacked into and from the nuclear core in a series of motions each involving moving the control rod a discrete incremental distance or "step"; hence, such movement is commonly referred to as stepping of the control rods. This magnetic jack type of mechanism is illustrated and described in U.S. Patents to Frisch (3,158,766) and DeWesse (3,992,255) which are assigned to the assignee of the present invention. This magnetic jack type of control rod drive mechanism includes three electromagnetic coils and armatures or plungers which are operated to raise and lower a drive rod shaft and thereby the control rod cluster assembly The three coils are mounted about and outside of the pressure housing. Two of the coils actuate respective plungers of movable and stationary grippers contained within the housing The third coil actuates a lift plunger connected to the movable gripper. Actuation of the movable and stationary plungers, in turn, operate sets of circumferentially spaced latches which grip the drive rod shaft having multiple axially-spaced circumferential grooves. The stationary gripper latches are actuated to hold the drive rod shaft in a desired axial position. The movable gripper latches are actuated to raise and lower the drive rod shaft. Each jacking or stepping movement of the control rod drive mechanism moves the drive rod shaft 5/8 inch. (1.58 cm). The jacking or stepping movement is thus accomplished by the operation of the three sets of axially spaced electromagnetic coils to actuate the corresponding stationary, movable and lift plungers so as to alternately and sequentially grip, move and release the control rod drive shaft of the respective mechanism. The construction and stepping mode of operation of the magnetic jack type of control rod drive mechanism as used heretofore have certain drawbacks. The drawback associated with the construction of the control rod drive mechanism derives from the pivotal mounting geometry of the latches of the stationary and movable grippers of the mechanism. These latches each have either one or a pair of teeth which engage within either a single groove or a pair of adjacent grooves in the drive rod shaft. Because the latches pivotally move in arcuate paths toward and away from drive rod shaft, the teeth are placed at locations on the latch body farthest from the pivotal axis of the latch. This latch teeth placement results in the generation of a moment load through the body of the latch which over time tends to cause cracking at the root of the teeth and eventual failure of the latch. Further, due to the limited number of teeth, the maximum being two, the effective wear life of the latch is limited. The drawback associated with the mode of operation of the control rod drive mechanism derives from the particular scheme employed in repositioning of the rod drive shaft and thereby each control rod cluster assembly during each nuclear reactor fuel cycle. During normal reactor operation, the control rod drive mechanisms hold the cluster assemblies of control rods withdrawn above the reactor core within guide tubes extending upwardly from and in alignment with the guide thimbles of the fuel assemblies. The force produced by upward flow of coolant causes a surface area of the each control rod wall to contact and rub against the inside of its associated guide tube. Eventually, the wall thickness of the control rod at the area of contact will decreased below allowable limits requiring replacement of the control rod. To avoid occurrence of rubbing contact wear at a single surface area of the control rod walls and thereby extend the useful life of the control rods, the drive rod shafts of the control rod drive mechanisms and thereby the control rods of the cluster assemblies are moved three steps at every refueling which occurs at the end of each fuel cycle of the nuclear reactor. Ideally, a one-step relocation is all that is needed to present a fresh surface area on the control rod to the portion of the guide tube where the wear is taking place. However, the inherent nature of the magnetic jack type of mechanism is that when it is actuated an occasional misstep will randomly occur in which the control rod is not moved. This presents an unacceptable degree of uncertainty for any given repositioning of the control rods as to whether or not the mechanism actually accomplished relocation of the control rods through one step if the cluster assemblies are only intended to be moved one step. If one-step relocation was not accomplished, or in other words a random misstep occurred, the additional exposure time of the original surface areas of the control rods for another complete fuel cycle at the same elevations would result in unacceptable wear at such areas. The unacceptable risk of a random misstep occurring is offset by the conventional practice of overcompensating and repositioning the control rods through a distance equivalent to three steps, instead of one step. Therefore, even if a one-step misstep occurs, the control rods will still be repositioned through two steps. The current strategy of repositioning the control rods in three-step increments handles the uncertainties of positioning in a simple way; by relocating the control rods an amount that is greater than any potential misstepping uncertainty. The drawback of the three-step repositioning scheme is that the wear is still poorly distributed throughout the total available clad thickness of the control rods. As an example, assume that a particular nuclear plant has a control rod wear rate that uses 60% of the permissible wear thickness in one fuel cycle. With the three-step repositioning scheme, the control rods would be relocated every fuel cycle so that the permissible minimum level of wear thickness is not exceeded. This effectively "wastes" 40% of the wear thickness at that elevation of the control rods. Consequently, a need exists for improvements in the construction and mode of operating the magnetic jack type of control rod drive mechanism employed in nuclear reactors so as to overcome the above-described drawbacks. SUMMARY OF THE INVENTION The present invention provides an improvement designed to satisfy the aforementioned needs. Particularly, the present invention is directed to a method of operating a control rod drive mechanism for performing single-step multiple repositionings of a control rod cluster assembly during a single fuel cycle of the nuclear reactor. The invention of the patent application cross-referenced above is directed to improvement of the prior art latch and linkage arrangement. In accordance with the present invention, a single step movement of the control rod cluster assembly is performed but on a more frequent basis than once every fuel cycle. With more frequent movement, an occasional misstep can still cause the control rods to remain at their previous elevation and continue wearing at the same surface area. But now the control rods will be repositioned more frequently, so the additional wear will be smaller than it would have been under the prior art once-per-fuel cycle repositioning scheme. By increasing the frequency of the repositioning to more than once per fuel cycle, the wear exposure time due to a misstep is decreased, and the consequences of a misstep are reduced proportionately. The more frequent repositioning better utilizes the control rod clad thickness available for wear. For example, if the control rod cluster assembly is repositioned three times per fuel cycle, the wear would be 20% of the allowable wear per repositioning. Eventually, each wear location could be used five times. With a wear of 20% per repositioning, 100% of the permissible wear thickness would be utilized. Accordingly, the present invention is directed to a method of repositioning a control rod cluster assembly of a nuclear reactor. The repositioning method comprises the steps of: (a) performing a single-step repositioning of a control rod cluster assembly in a nuclear reactor; and (b) repeating the single-step repositioning at a plurality of separate times during a given fuel cycle of the reactor to perform a sequence of at least more than one single-step repositioning of the control rod cluster assembly at separate spaced times during the given fuel cycle of the nuclear reactor. In one embodiment of the method, the singlestep repositioning is repeated once every four months in a twelve month fuel cycle. In another embodiment, the single-step repositioning is repeated once every month in a twelve month fuel cycle. These and other features and advantages of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
050292498 | claims | 1. An electron microscope comprising: a hollow sample chamber; evacuation means communicating with the interior of the sample chamber for evacuating the sample chamber; a casing mounted on said sample chamber; an electron lens system provided within the casing; an electron gun chamber mounted on the casing; and an electron beam generating means provided in the electron gun chamber for generating an electron beam; wherein an electron beam path is defined from the electron beam generating means to the sample chamber, and wherein the electron gun chamber, the casing, and the sample chamber define a closed space and communicate with each other so as to define an evacuation path within said closed space for enabling evacuation of said electron gun chamber by said evacuation means, said evacuation path being independent of said electron beam path. a hollow sample chamber; evacuation means communicating with the interior of the sample chamber for evacuating the sample chamber; a casting mounted on said sample chamber; an electron lens system provided within the casing; an electron gun chamber mounted on the casing; and an electron beam generating means provided in the electron gun chamber for generating an electron beam; wherein an electron beam path is defined from the electron beam generating means to the sample chamber, and wherein an evacuation path is defined within the casing so as to extend between the sample chamber and the electron gun chamber for enabling evacuation of the electron gun chamber by said evacuation means, said evacuation path being offset relative to the electron beam path. a hollow sample chamber; evacuation means communicating with the interior of the sample chamber for evacuating the sample chamber; a casing mounted on said sample chamber; an electron lens system provided in the casing; an electron gun chamber mounted on the casing; and an electron beam generating means provided in the electron gun chamber for generating an electron beam; wherein an electron beam path is defined from the electron beam generating means to the sample chamber, and wherein an evacuation path is defined within the casing for enabling evacuation of the casing via the sample chamber, said evacuation path being offset relative to the electron beam path. said electron microscope being characterised by the electron gun chamber, the casing, and the sample chamber defining a closed space communicating with the evacuating means, and an evacuation path being defined between the electron gun chamber, the casing, and the sample chamber, said evacuation path being different from a path of said electron beam. a hollow sample chamber; a casing; a hollow electron gun chamber; and a plurality of sealed modules provided within the casing, each of said sealed modules containing an electron lens and having a bore therethrough for defining an electron beam path; wherein a communication path is defined in the casing from the sample chamber to the electron gun chamber around said sealed modules, at least part of that path being defined between an outer wall of each of said sealed modules and an inner wall of said casing. a hollow casing; a plurality of modules provided within said casing, each of said modules having at least one electron lens sealed therein; a hollow electron gun chamber secured to one end of said casing; and means for delimiting at least one evacuation bore extending to the interior of the electron gun chamber; wherein at least part of an outer wall of each of said modules is spaced from an inner wall of the casing so as to define part of an evacuation path around each of said modules, and wherein the evacuation path extends from said electron gun chamber, through said at least one evacuation bore, and around said modules, to an end of said casing remote from said electron gun chamber. a hollow sample chamber; a hollow casing; a tie member interconnecting the sample chamber and the casing, the tie member having plurality of bores provided therein; an electron lens system provided within the casing, at least one gas flow path being provided within said casing around said electron lens system; a hollow electron gun chamber; and a spaced interconnecting the casing and the electron gun chamber, the spacer having a plurality of through-bores provided therein. 2. An electron microscope according to claim 1, wherein the electron lens system comprises a plurality of sealed modules, each of said modules containing an electron lens and having a bore therethrough for defining part of said electron beam path. 3. An electron microscope according to claim 2, having a spacer between said modules, said spacer having an opening therein, said opening forming part of said evacuation path. 4. An electron microscope according to claim 2, wherein at least part of an outer wall of each of said modules is spaced from an inner wall of the casing, that spacing forming part of said evacuation path. 5. An electron microscope according to claim 1, further comprising a spacer between said electron gun chamber and said casing, said spacer having a plurality of through-bores provided therein, said through-bores defining part of said evacuation path. 6. An electron microscope according to claim 1, further comprising a tie member between said casing and said sample chamber, said tie member having a plurality of bores provided therein, said bores defining part of said evacuation path. 7. An electron microscope according to claim 1, further comprising magnetic shielding enclosure surrounding at least the casing. 8. An electron microscope according to claim 7, wherein said shielding enclosure is sealed to said sample chamber. 9. An electron microscope according to claim 7, wherein said shielding enclosure also surrounds said electron gun chamber. 10. An electron microscope comprising: 11. An electron microscope comprising: 12. An electron microscope comprising a hollow electron gun chamber in which an electron beam generating means for generating an electron beam is provided, a hollow casing for containing an electron lens system, a hollow sample chamber, and evacuating means for evacuating the sample chamber; 13. An electron microscope comprising: 14. An electron beam column for an electron microscope, comprising; 15. An electron microscope comprising: 16. An electron microscope according to claim 15, wherein said electron lens system defines an electron beam path separate from said gas flow path. |
052689484 | claims | 1. A locking assembly for permitting the easy removal of fuel rods from a nuclear power reactor fuel bundle, said locking assembly including: (a) a plurality of guide tubes extending upwardly from the fuel bundle; (b) a support plate removably mounted on said guide tubes to permit removal of said fuel rods when said support plate is removed from said guide tubes, said support plate being formed with an opening and being formed with a slot extending outwardly from said opening; (c) a collar assembly mounted at the upper extending end of each said guide tube for permitting selective removal of said support plate from said guide tubes, said collar assembly including a base portion and first and second locking portions mounted on said base portion for movement relative to one another, one of said locking portions having a projection formed thereon corresponding generally in shape to said support plate slot, and one of said locking portions being movable between a first position at which said projection is openly aligned with said slot to permit removal of said support plate and a second position at which said projection is not openly aligned with said slot whereby said plate is locked in place on said guide tubes; and (d) resilient means associated with said collar assembly, said resilient means being located on one of said locking portions for resiliently engaging a surface of the collar assembly with sufficient force to maintain said movable locking portion at said second locking position thereof, and for releasing said engaged surface when a predetermined torsional force is applied to said movable locking portion, thereby permitting movement of said movable locking portion to said first aligned position thereof at which said support plate can be removed from said guide tubes. (a) a plurality of guide tubes extending upwardly from the fuel bundle; (b) a support plate removably mounted on said guide tubes to permit removal of said fuel rods when said support plate is removed from said guide tubes, said support plate being formed with an opening and being formed with a slot extending outwardly from said opening; (c) a collar assembly mounted at said extending end of said guide tube for permitting selective removal of said support plate from said guide tubes, said collar assembly including a base portion, a first locking portion arranged for rotational movement with respect to said base portion and having a body portion formed with a slot which corresponds in shape to said slot in said support plate, and a second locking portion fixed to said base portion and including a projection extending outwardly therefrom, said first locking portion being disposed intermediate said projection and said support plate and being rotatable between a first position at which said slot therein is openly aligned with said slot in said support plate to permit said support plate to pass over said projection for removal, and a second position at which said slot therein is out of alignment with said slot in said support plate and said body portion is disposed between said projection and said support plate to prevent removal of said support plate over said projection; and (d) resilient means for resiliently maintaining said first locking portion at said second position thereof and for permitting movement of said first locking portion to said first position thereof when a predetermined torsional force is applied to said first locking portion. (a) a plurality of guide tubes extending upwardly from the fuel bundle; (b) a support plate removably mounted o said guide tubes to permit removal of said fuel rods when said support plate is removed from said guide tubes, said support plate being formed with an opening and being formed with a slot extending outwardly from said opening; (c) a collar assembly mounted at said extending end of each said guide tube for permitting selective removal of said support plate from said guide tubes, said collar assembly including a base portion fixed to said guide tube, and a locking portion rotationally mounted o said base portion and including a projection corresponding generally in shape to said slot in said support plate, said locking portion being rotatable between a first position at which said projection is openly aligned with said support plate slot to permit removal of said support plate and a second position at which said projection overlies said support plate to prevent removal thereof, and said locking portion being formed with an interior annular surface; and (d) resilient means associated with said collar assembly for resiliently locking said locking portion against movement when it is at said second position thereof, said resilient means including a spring biased element disposed within and adjacent said interior engagement surface of said locking portion for resiliently engaging said fixed base portion with a sufficient force to normally maintain said locking portion at said first position thereof and for releasing said locking portion when a predetermined torsional force is applied thereto. (a) a plurality of guide tubes extending upwardly from the fuel bundle; (b) a support plate removably mounted on said guide tubes to permit removal of said fuel rods when said support plate is removed from said guide tubes, said support plate being formed with an opening and being formed with a slot extending outwardly from said opening; (c) a collar assembly mounted at the extending end of each said guide tube for permitting selective removal of said support plate from said guide tubes, said collar assembly including a base portion, a movable locking portion rotatably mounted on said base portion and being formed with a projection corresponding generally in shape to said support plate slot, and a retaining member fixed to said guide tube maintaining said movable locking portion in place on said base portion, said movable locking portion being movable between a first position at which said projection is aligned with said support plate slot to permit removal of said support plate and a second position at which said projection overlies said support plate to prevent removal thereof; and (d) resilient means for resiliently maintaining said movable locking portion at said second position thereof and for permitting movement of said movable locking portion to said first position thereof when a predetermined torsional force is applied thereto, said resilient means including cooperating portions of both said movable locking portion and said retaining member. 2. A locking assembly as defined in claim 1 wherein one of said locking portions includes an annular body portion presenting an annular engagement surface, and said resilient means resiliently locks said movable locking portion against movement by engagement at said annular engagement surface. 3. A locking assembly as defined in claim 2 wherein said resilient means includes a detent formed in said annular engagement surface of said movable locking portion, and a spring biased ear resiliently engaging said detent at said second locking position of said movable locking portion and disengaging said detent at said first position of said movable locking portion. 4. A locking assembly as defined in claim 3 wherein said resilient means includes a second detent formed in said annular engagement surface of said movable locking portion, and said spring biased ear engages said second detent at said first position of said movable locking portion. 5. A locking assembly as defined in claim 2 wherein said resilient means includes a fixed annular spring biased element disposed in slidable abutting relation with said annular engagement surface of said movable locking portion. 6. A locking assembly as defined in claim 5 wherein said annular spring biased element receives within its confines said movable locking portion, said annular engagement surface of said movable locking portion includes at least one detent formed therein, and said annular spring biased element includes a spring biased ear for resiliently engaging said detent at said second locking position of said movable locking portion. 7. A locking assembly as defined in claim 2 wherein said annular engagement surface is located at the interior face of said movable locking portion and includes a locking protrusion extending inwardly therefrom, and said resilient means includes a spring biased element received within said movable locking portion and being formed with an opening for receiving said inwardly extending protrusion for causing said spring biased element to move with said movable locking portion. 8. A locking assembly as defined in claim 2 wherein said first locking portion is said movable locking portion and is formed with a spring biased ear in said annular engagement surface, and said second locking portion is received within said first locking portion and is formed with at least one detent resiliently engaged by said spring biased ear at said second locking position of said first locking portion. 9. A locking assembly as defined in claim 2 wherein said first locking portion is said movable locking portion, and said second locking portion is fixed to said base portion and is formed with said projection, with said first locking portion being located intermediate said projection and said support plate and being formed with a second slot corresponding to said slot in said support plate, whereby said support plate can be removed when said first locking portion is moved to said first position at which said slot in said first locking portion is aligned with said slot in said support plate, and whereby said first locking portion can be moved to said second position at which said respective slots are not aligned and said support plate is locked against removal by said first locking portion being disposed between said projection and said support plate. 10. A locking assembly for permitting the easy removal of fuel rods from a nuclear power reactor fuel bundle, said locking assembly including: 11. A locking assembly as defined in claim 10 wherein said first locking portion includes an exterior annular engagement surface, and said resilient means also resiliently engages said annular engagement surface at said first position of said first locking portion. 12. A locking assembly as defined in claim 11 wherein said annular engagement surface includes at least two detents located at spaced location therein, and wherein said resilient means includes a spring biased ear resiliently engaging one of said detents when said first locking portion is at said first position thereof and resiliently engaging the other of said detents when said first locking portion is at said second position thereof. 13. A locking assembly as defined in claim 12 wherein said spring biased ear is formed at the interior annular surface of a locking cup member that receives said first locking portion and positions said exterior annular surface thereof adjacent said interior annular surface of said locking cup member. 14. A locking assembly for permitting the easy removal of fuel rods from a nuclear power reactor fuel bundle, said locking assembly including: 15. A locking assembly as defined in claim 14 wherein said annular surface of said locking portion has an inwardly extending protrusion and said spring biased element is a snap ring positioned within said locking portion for rotational movement therewith and formed with indentations for resiliently engaging said fixed base portion. 16. A locking assembly for permitting the easy removal of fuel rods from a nuclear power reactor fuel bundle, said locking assembly including: 17. A locking assembly as defined in claim 16 wherein said movable locking member includes an open interior annular wall portion, said retaining member includes an exterior annular wall portion disposed within and adjacent said open interior annular wall portion, and said resilient means includes a detent formed in said exterior annular wall portion and a spring biased ear formed in said interior annular wall portion, whereby said movable locking portion is normally resiliently maintained at said second position thereof when said spring biased ear is within said detent. 18. A locking assembly as defined in claim 17 wherein said interior annular wall of said movable locking portion is formed of deflectable metal and includes an indentation that is normally deflected inwardly to form said spring biased ear and that can be deflected outwardly when said predetermined torsional force is applied to said movable locking portion to move it from said second position to said first position. 19. A locking assembly as defined in claim 18 wherein said retaining member includes a radially projecting shoulder that extends over said interior annular wall portion of said movable locking portion to maintain it in place on said base portion, said shoulder being formed with at least one axial groove positioned therein to permit a tool to be pushed therethrough to form said spring biased ear in said movable locking member. |
050935798 | abstract | A substrate holding device includes a holding table having a reduced pressure passageway; a pressure gauge for measuring a value related to the pressure in the reduced pressure passageway; a pump for producing a pressure difference between a first surface of the substrate to be attracted to the holding table and a second surface of the substrate not to be attracted to the holding table; a valve which can be opened/closed for control of the pressure in the reduced pressure passageway; a pressure control system for controlling the opening/closing of the valve on the basis of an output corresponding to the value measured by the pressure gauge; and a temperature control system for controlling the temperature of the holding table. |
description | This application is a Continuation Application of the commonly-owned U.S. patent application with Ser. No. 11/712,640, now U.S. Pat. No. 7,397,901, filed Feb. 28, 2007, by S. Johnsen, and entitled “A Multi-Leaf Collimator with Leaves Formed of Different Materials.” Embodiments of the invention relate generally to radiation shielding. More specifically, embodiments of the invention pertain to a multi-leaf collimator that can be used in applications such as radiotherapy. Multi-leaf collimators (MLC) are commonly used in radiotherapy machines. Typically, an MLC includes two sets of independently adjustable leaves. Each leaf is thick enough to attenuate or block completely a beam of radiation. The first set of leaves is positioned on one side of the beam's path, and the second set mirrors the first set on the other side of the beam's path. The leaves can be positioned independently of one another to form an aperture in a shape like that of the area to be irradiated, so that only the targeted area is irradiated while surrounding areas are shielded. Because adjacent leaves need to be able to move relative to one another, there is necessarily a gap between them. The leaves are designed so that one or more portions of one leaf overlap one or more corresponding portions of an adjacent leaf. Thus, the gap between adjacent leaves does not provide a straight-line passageway for the radiation beam. Instead, radiation that enters a gap will encounter a thickness of leaf material that is sufficient for attenuating or blocking the radiation. In this manner, leakage of radiation through the gaps to areas other than the targeted area is prevented. Conventional MLC leaves may be formed using only a high density material or a material with a high atomic number (referred to as a high-Z material). High density/high-Z material is more effective at blocking a radiation beam than low density/low-Z material. However, for proton beams, the use of high density/high-Z material in an MLC can result in the production of more neutrons than the use of low density/low-Z material. Additional neutrons can be undesirable because they contribute to the total dose received by a patient. High density/high-Z material can also be difficult to machine and relatively expensive, and leaves made from such material are heavier and hence more difficult to move than leaves made of low density/low-Z material. On the other hand, conventional leaves made only of low density/low-Z material are thicker—perhaps substantially thicker—than leaves made of high density/high-Z material. According to an embodiment of the present invention, an MLC includes a number of independently adjustable and overlapping leaves. The overlapping portions of the leaves are made of a first material, while the non-overlapping portions of the leaves are made of a second, different material. The first material is denser (or has a higher atomic number) than the second material. Accordingly, the majority of each leaf is made using the second (e.g., less dense) material, while a smaller portion of each leaf is made from the first (e.g., denser) material. Consequently, leaves made using a combination of low density (or low-Z) and high density (or high-Z) materials can be lighter, less expensive, easier to fabricate and will produce less neutrons than conventional leaves made using only a high density or high-Z material, and thinner than conventional leaves made using only a low density or low-Z material. Unless noted otherwise, the drawings are not to scale. Reference will now be made in detail to embodiments of the present invention. While the invention will be described in conjunction with these embodiments, it will be understood that they are not intended to limit the invention to these embodiments. On the contrary, the invention is intended to cover alternatives, modifications and equivalents, which may be included within the spirit and scope of the invention as defined by the appended claims. Furthermore, in the following detailed description of the present invention, numerous specific details are set forth in order to provide a thorough understanding of the present invention. However, it will be recognized by one of ordinary skill in the art that the present invention may be practiced without these specific details. In other instances, well known methods, procedures, and components have not been described in detail as not to unnecessarily obscure aspects of the present invention. FIG. 1 is a perspective view of a portion of a multi-leaf collimator (MLC) 100 in accordance with one embodiment of the present invention. Generally speaking, MLC 100 is designed to define and restrict (constrict) the dimensions of a radiation beam 122. In one embodiment, MLC 100 is implemented as part of a radiotherapy machine. In one embodiment, radiation beam 122 is a beam of x-rays, and in another embodiment, radiation beam 122 is a beam of charged particles such as protons, electrons or heavy ions. In the example of FIG. 1, MLC 100 includes a set of plates or leaves, exemplified by leaves 110, 111, 112 and 113. Leaves 111 and 112 are also isolated in cross-sectional view. MLC 100 may include other components not illustrated or discussed herein, such as guides or tracks for the leaves, drive mechanisms, suspension mechanisms, and so on. As illustrated in FIG. 2, there may be a second set of leaves opposing the first set in a symmetrical or mirror-like fashion. In the examples of FIGS. 1 and 2, any leaf can be adjusted independently of any other leaf. There is a gap (e.g., gap 140) between adjacent leaves (e.g., leaves 111 and 112) so that the leaves can move relative to one another. Leaves can be moved back and forth in the directions indicated by the arrows 115 and 215. With reference to FIG. 1, the MLC 100 is situated between a source 120 of the radiation beam 122 and a target area 130. The shape of the target area 130—that is, the shape of the area to be exposed to the beam 122—is defined by positioning the leaves of the MLC 100 accordingly. The leaves block portions of the beam 122, while other portions of the beam reach the defined target area 130 unblocked. Unless otherwise indicated, the word “block” is used herein in the general sense to mean either “prevent” (completely block) or “hinder” (partially block or attenuate). Thus, areas outside of the target area 130 are shielded at least to some extent from the beam 122, so that areas outside the target area are not exposed to unnecessary levels of radiation. The leaves 111 and 112 are designed with overlapping portions 151 and 152, respectively, which are situated between adjacent surfaces of those leaves. In the discussion herein, an overlapping portion 151 or 152 of a leaf may be referred to as the “first portion” of a leaf, while the portion of a leaf other than the overlapping portion (e.g., portion 161 or 162) may be referred to as the “second portion” or the “non-overlapping portion” of the leaf. The overlapping portions 151 and 152 can also be described as being complementary. In the example of FIG. 1, leaf 111 also includes an overlapping portion 153 that is situated between adjacent surfaces of leaf 111 and adjacent leaf 110, and leaf 112 also includes an overlapping portion 154 that is situated between leaf 112 and an adjacent leaf 113. Each leaf in MLC 100 may be similarly designed. The portions 161 and 162 each have a width (measured in the direction W of FIG. 1) and a height or thickness (measured in the direction H of FIG. 1). In a radiotherapy machine that uses x-rays, for example, the x-rays will be attenuated as they pass through a portion 161 or 162, in which case the portions 161 and 162 can each be made thick enough to reduce the x-ray dosage by a desired amount. In contrast to x-rays, charged particles such as protons and heavy ions are not attenuated in number but have a definite range in a given material. Accordingly, in a radiotherapy machine that uses charged particles, the portions 161 and 162 can each be made thick enough to completely block the charged particles. That is, the thickness of material used to form each of the portions 161 and 162 meets or exceeds the range of charged particles in the material. Also, in a radiotherapy machine that uses x-rays, the overlapping portions 151-154 can each be made thick enough to reduce the x-ray dosage by the desired amount. That is, overlapping portion 151 can by itself attenuate an x-ray by the desired amount, and overlapping portion 152 can also by itself attenuate an x-ray that enters gap 140 (bypassing overlapping portion 151) by the desired amount. Similarly, in a radiotherapy machine that uses charged particles, the overlapping portions 151-154 can each be made thick enough to completely block the charged particles. Thus, overlapping portion 151 can by itself completely block charged particles, and overlapping portion 152 can also by itself completely block charged particles that enter gap 140 (bypassing overlapping portion 151). According to embodiments of the present invention, the overlapping portions 151-154 are made of a material that is different from the material used to make portions 161 and 162. In one embodiment, the overlapping portions 151-154 are made of a material (which may be referred to herein as the “first material”) that has a higher density than the material (which may be referred to herein as the “second material”) that makes up portions 161 and 162. In another embodiment, the overlapping portions 151-154 are made of a material that has a higher atomic number (a higher-Z material) than the material (a lower-Z material) that makes up portions 161 and 162. Higher density or higher-Z materials (“higher density/higher-Z materials”) include, but are not limited to, tungsten, tungsten alloys, tantalum, tantalum alloys, lead or lead alloys, while lower density or lower-Z materials (“lower density/lower-Z materials”) include, but are not limited to, steel (various steel alloys), brass, zinc or copper. Generally speaking, a higher density/higher-Z material may be a material that has a density of about 15 gm/cm3 or greater. In one embodiment, the width of an overlapping portion (e.g., portion 151) is approximately 0.25-0.5 mm, while the width of a non-overlapping portion is approximately 0.5-1 cm. In one embodiment, the height of an overlapping portion is approximately 3 cm while the height of a non-overlapping portion is approximately 6 cm. In general, for each leaf of MLC 100, the volume of the non-overlapping portion is greater than the total volume of the overlapping portions. Thus, according to embodiments of the present invention, the bulk of each leaf in MLC 100 is made using a lower density/lower-Z material, while a lesser portion of each leaf is made using a higher density/higher-Z material. Leaves formed of different materials in this manner provide a number of advantages compared to leaves formed only of either higher density/higher-Z material or lower density/lower-Z material. For one, because the leaves of MLC 100 utilize a reduced amount of higher density/higher-Z material relative to conventional leaves formed only of higher density/higher-Z material, they will weigh less than conventional leaves. Because they are lighter, the leaves of MLC 100 are easier to move and so the mechanisms for positioning them can be less robust. Also, higher density/higher-Z material can be difficult to work with (e.g., machine), and so by reducing the amount of such material, the leaves of MLC 100 can be easier to fabricate than conventional leaves. In addition, higher density/higher-Z material can be more expensive, and so by reducing the amount of such material, the leaves of MLC 100 can cost less than conventional leaves. Furthermore, higher density/higher-Z material produces more neutrons when subject to a proton beam or x-ray beam, and so by reducing the amount of such material, the neutron dose to a patient, for example, can be reduced. On the other hand, because the leaves of MLC 100 still utilize higher density/higher-Z material in the overlapping portions of the leaves, the leaves of MLC 100 can be made thinner than conventional leaves formed only of lower density/lower-Z material. For example, the range of 250 MeV protons in brass is about 65 mm while their range in tungsten is about 40 mm. Because each of the overlapping portions 151-154 is thick enough by itself to block radiation beam 122, at a minimum the thicknesses of the portions 151-154 are equal. Therefore, at a minimum the thicknesses of the leaf portions 161 and 162 are each about twice the height of a single overlapping portion 151, 152, 153 or 154. Thus, in conventional leaves made only of brass, the minimum thickness of an overlapping portion 151-154 would be about 65 mm and the minimum thickness of a non-overlapping portion 161-162 would be about 130 mm. However, according to embodiments of the present invention, the overlapping portions (e.g., portions 151 and 153) of a leaf (e.g., leaf 111) may be made of tungsten, while the remaining portion (e.g., portion 161) of the leaf may be made of brass. If the thicknesses of the overlapping portions made of tungsten are to exceed the range of protons in tungsten, the minimum thickness of an overlapping portion 151-154 would be about 40 mm. Accordingly, the minimum thickness of a non-overlapping portion 161-162 would be about 80 mm, representing a significant reduction in the thickness of MLC 100 relative to a conventional MLC. In general, fabrication of the leaves in MLC 100 (e.g., leaves 111 and 112) entails attaching or bonding the higher density/higher-Z portion and the lower density/lower-Z portion. This can be achieved in a number of different ways, depending on the materials used. The portions may be joined using a technique such as brazing, for example. Alternatively, the portions may be joined mechanically (e.g., using screws or other types of fasteners). FIG. 2 is a top-down view of a portion of an MLC 200 in accordance with one embodiment of the present invention. MLC 200 includes a number of leaves exemplified by leaves 210 and 211, which are also shown in cross-section along axis A-A. The leaves can be independently moved back and forth in the direction of arrow 215, in order to create an aperture 220 that defines the shape of a targeted area. Generally speaking, the leaves 210 and 211 lie on a common axis and move back and forth along that axis. There is a gap 225 between opposing leaves (e.g. leaves 210 and 211) when those leaves are in the closed position. Continuing with reference to FIG. 2, the leaves 210 and 211 are designed with overlapping portions 230 and 231, respectively, which are situated between adjacent surfaces of those leaves. Each leaf in MLC 200 may be similarly designed. As before, an overlapping portion 230 or 231 may be referred to as the “first portion” of a leaf, while the portion of a leaf other than the overlapping portion (e.g., portion 240 or 241) may be referred to as the “second portion” or “non-overlapping portion” of the leaf. The overlapping portions 230 and 231 can also be described as complementary. The non-overlapping portions 240 and 241 each have a height or thickness (measured in the direction H of FIG. 2). In a radiotherapy machine that uses x-rays, for example, the x-rays will be attenuated as they pass through a portion 240 or 241, in which case the portions 240 and 241 can each be made thick enough to reduce the x-ray dosage by a desired amount. In a radiotherapy machine that uses charged particles, the portions 240 and 241 can each be made thick enough to completely block the charged particles. That is, the thickness of each of the portions 240 and 241 can be chosen to meet or exceed the range of the charged particles in the material used to form those portions. Also, in a radiotherapy machine that uses x-rays, the overlapping portions 230 and 231 can each be made thick enough to reduce the x-ray dosage by the desired amount. That is, overlapping portion 230 can by itself attenuate an x-ray by the desired amount, and overlapping portion 231 can also by itself attenuate an x-ray that enters gap 225 (bypassing overlapping portion 230) by the desired amount. Similarly, in a radiotherapy machine that uses charged particles, the overlapping portions 230 and 231 can each be made thick enough to completely block the charged particles. Thus, overlapping portion 230 can by itself completely block charged particles, and overlapping portion 231 can also by itself completely block charged particles that enter gap 225 (bypassing overlapping portion 230). According to embodiments of the present invention, the overlapping portions 230 and 231 are made of a material that is different from the material used to make non-overlapping portions 240 and 241. In one embodiment, the overlapping portions 230 and 231 are made of a material (which may be referred to herein as the “first material”) that has a higher density than the material (which may be referred to herein as the “second material”) that makes up portions 240 and 241. In another embodiment, the overlapping portions 230 and 231 are made of a higher-Z material than the material that makes up portions 240 and 241. Thus, according to embodiments of the present invention, the bulk of each leaf in MLC 200 (that is, the second or non-overlapping portion of each leaf) is made from a less dense/lower-Z material than the overlapping (first) portions of each leaf. Leaves formed of different materials in this manner provide a number of advantages compared to leaves formed entirely of either higher density/higher-Z material or lower density/lower-Z material. Those advantages have been previously discussed herein. Furthermore, the embodiments discussed in conjunction with FIG. 1 and the embodiments discussed in conjunction with FIG. 2 can be combined. That is, overlapping portions can be situated on adjacent surfaces of parallel leaves and on adjacent ends of opposing leaves. FIG. 3 is a cross-sectional view of adjacent leaves 310 and 311 according to an embodiment of the present invention. Leaf 310 includes overlapping portions 320 and 321 and non-overlapping portion 330, and leaf 311 includes overlapping portions 322 and 323 and non-overlapping portion 331. The overlapping portions 320-323 are made of a material that is different from the material used to make portions 330 and 331. In one embodiment, the overlapping portions 320-323 are made of a material that has a higher density than the material that makes up portions 330 and 331. In another embodiment, the overlapping portions 320-323 are made of a higher-Z material than the material that makes up portions 330 and 331. In the example of FIG. 3, the leaves 310 and 311 include structural elements 350 and 351, which represent features of the leaves such as drive mechanisms or suspension mechanisms. In one embodiment, the structural elements 350 and 351 are made of the same lower density/lower-Z material as the leaf portions 350 and 351. In one such embodiment, the structural elements 350 and 351 are machined from the same block of material used to form the leaf portions 330 and 331. As mentioned above, lower density/lower-Z material is easier to work with than higher density/higher-Z material. Thus, another advantage associated with forming the non-overlapping leaf portions from lower density/lower-Z material is that the structural elements 350 and 351 can be easier to machine. Leaf elements 320-323 may rely on the other parts of the leaves, such as elements 330 and 331, to provide structural strength. Therefore, the elements 320-323 may be made from material with lesser strength, such as lead or its alloys, than the material that makes up the other portions of the leaves. This provides more flexibility in the design of the leaves. FIGS. 4 and 5 illustrate cross-sectional views of adjacent leaves according to other embodiments of the present invention. In the example of FIG. 4, adjacent leaves 410 and 411 are shown. Leaf 410 includes non-overlapping portion 430 and overlapping portions 420, 421 and 422 on the surface adjacent to leaf 411, and leaf 411 includes non-overlapping portion 431 and overlapping portions 423, 424 and 425 on the surface adjacent to leaf 410. The overlapping portions 420-422 may be collectively referred to as the “first portion” of leaf 410 and the overlapping portions 423-425 may be collectively referred to as the “first portion” of leaf 411. The overlapping portions 420-425 are made of a material that is different from the material used to make portions 430 and 431. In one embodiment, the overlapping portions 420-425 are made of a material that has a higher density than the material that makes up portions 430 and 431. In another embodiment, the overlapping portions 420-425 are made of a higher-Z material than the material that makes up portions 430 and 431. In general, the total thicknesses (H) of the overlapping portions 420-422, and of the overlapping portions 423-425, are sufficient to attenuate or completely block a radiation beam. In the example of FIG. 5, adjacent leaves 510 and 511 are shown. Leaf 510 includes non-overlapping portion 530 and overlapping portions 520 and 521 on the surface adjacent to leaf 511, and leaf 511 includes non-overlapping portion 531 and overlapping portion 523 on the surface adjacent to leaf 510. The overlapping portions 520 and 521 may be collectively referred to as the “first portion” of leaf 510. The overlapping portions 520, 521 and 523 are made of a material that is different from the material used to make portions 530 and 531. In one embodiment, the overlapping portions 520, 521 and 523 are made of a material that has a higher density than the material that makes up portions 530 and 531. In another embodiment, the overlapping portions 520, 521 and 523 are made of a higher-Z material than the material that makes up portions 530 and 531. In general, the total thicknesses (H) of the overlapping portions 520 and 521, and of the overlapping portion 523, are sufficient to attenuate or completely block a radiation beam. FIG. 6 illustrates a cross-sectional view of adjacent leaves 610 and 611 according to another embodiment of the present invention, referred to herein as a sawtooth embodiment. Leaf 610 includes non-overlapping portion 630 and a number of overlapping portions exemplified by portion 620, and leaf 611 includes non-overlapping portion 631 and a number of overlapping portions exemplified by portion 621. The overlapping portions 620 and 621 are made of a material that is different from the material used to make portions 630 and 631. In one embodiment, the overlapping portions 620 and 621 are made of a material that has a higher density than the material that makes up portions 630 and 631. In another embodiment, the overlapping portions 620 and 621 are made of a higher-Z material than the material that makes up portions 630 and 631. The thickness (H) of each leaf 610 and 611 in the sawtooth embodiment depends on how closely the overlapping portions of adjacent leaves interleave. The thickness of each sawtooth (e.g., portion 620 or 621) changes with its width (W). If the overlapping portions are far enough apart, an x-ray or charged particle may encounter only the thinner portions of each sawtooth. Accordingly, leaf thickness will increase in proportion to the number and widths of the gaps traversed by an x-ray or charged particle. The overlapping portions described above in conjunction with FIGS. 4, 5 and 6 can be situated on adjacent surfaces of parallel leaves and/or on adjacent ends of opposing leaves as in the example of FIG. 2. FIG. 7 is a flowchart 700 of a method of shaping a beam of radiation according to an embodiment of the present invention. Although specific steps are disclosed in flowchart 700, such steps are exemplary. That is, the present invention is well-suited to performing various other steps or variations of the steps recited in flowchart 700. In step 710, a first part of the beam is block (partially or completely, depending on the implementation) using a first thickness of a first material comprising a first portion of a first leaf in a multi-leaf collimator. In step 720, a second part of the beam is blocked (partially or completely, depending on the implantation) using a second thickness of a second material comprising a second portion of the first leaf. The first portion extends from a surface of the second portion to prevent the first part of the beam from passing through a gap between the first leaf and a second leaf in the multi-leaf collimator. The second material is different from the first material and the first thickness is less than the second thickness. In summary, according to embodiments of the present invention, the leaves of an MLC can be fabricated from a combination of materials. Specifically, a higher density/higher-Z material can be used in those portions of each leaf that overlap a corresponding portion of an adjacent leaf. The bulk of each leaf is made from a lower density/lower-Z material. Such leaves can be lighter, less expensive, easier to fabricate and will produce less neutrons or x-rays than conventional leaves made using only a high density or high-Z material, and thinner than conventional leaves made using only a low density or low-Z material. The foregoing descriptions of specific embodiments of the present invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise forms disclosed, and many modifications and variations are possible in light of the above teaching. The embodiments described herein were chosen and described in order to best explain the principles of the invention and its practical application, to thereby enable others skilled in the art to best utilize the invention and various embodiments with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto and their equivalents. |
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046817263 | abstract | A fast breeder reactor has a nuclear reactor container filled with a liquid metal, a reactor core disposed within the nuclear reactor container, and a first supporting structural member which is mounted to the nuclear reactor container such as to support the reactor core. The fast breeder reactor is provided with a cylindrical structural member which surrounds the periphery of the reactor core such as to define an annular gap between the cylindrical structural member and the reactor core for allowing the liquid metal to exist therein, the cylindrical structural member being mounted to the nuclear reactor container by means of a second supporting structural member. The inertial resisting force produced when the liquid metal existing in the annular gap will flow out from the annular gap in response to the vibration of the reactor core acts such as to suppress the vibration of the reactor core, thereby allowing an improvement in the anti-vibration properties of the fast breeder reactor. |
claims | 1. A plasma confinement system comprising:a confinement chamber comprising one or more ports located at an outer radius of the confinement chamber;one or more enclosures aligned substantially parallel to the outer radius and each coupled to a respective first port of the one or more ports at a first end of the enclosure and coupled to a respective second port of the one or more ports at a second end of the enclosure;a first set of one or more conductive coils respectively located between each of the one or more enclosures and the confinement chamber, wherein the first set of one or more conductive coils are aligned substantially parallel with the one or more enclosures; anda second set of one or more conductive coils respectively surrounding a portion of each of the one or more enclosures. 2. The plasma confinement system of claim 1, further comprising at least one conductor comprising a first conductor that is (i) external to the confinement chamber, (ii) aligned substantially parallel to the outer radius, and (iii) centered upon a vertical axis of the confinement chamber that is perpendicular to the outer radius. 3. The plasma confinement system of claim 2, wherein the first conductor is located on a plane containing the outer radius. 4. The plasma confinement system of claim 2, wherein the first conductor comprises:a first portion located adjacently above the one or more enclosures; anda second portion located adjacently below the one or more enclosures. 5. The plasma confinement system of claim 2, further comprising a neutron shield located between the one or more enclosures and the at least one conductor. 6. The plasma confinement system of claim 2, wherein the at least one conductor comprises a superconducting material. 7. The plasma confinement system of claim 1, further comprising:at least one sensor positioned on an inner wall of the confinement chamber; anda controller electrically coupled to the at least one sensor. 8. The plasma confinement system of claim 1, wherein the one or more enclosures comprise a plurality of enclosures, wherein each enclosure of the plurality of enclosures are (i) coupled to the confinement chamber at the outer radius and (ii) spaced symmetrically around the outer radius. |
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summary | ||
abstract | A containment airlock, in particular an airlock for intervention on a site likely to be contaminated with radiation, asbestos and biological and/or chemical agents. The containment airlock includes a self-supporting frame and a flexible containment shell. The shell is configured to be assembled with the frame. The frame is articulated so as to be extensible between a folded storage position and an extended intervention position. The frame includes articulated reinforcement rods. The articulated reinforcement rods include rigid segments and at least one intermediate articulation connecting the segments. |
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claims | 1. A fuel assembly handling tool comprising:a bail configured to be connected to a crane or other hoist;a bail plate connected to the bail as to be freely supported by the bail;a tool body freely supported at an upper end from the bail plate and extending between the bail plate and a lower end with the length between the upper and lower ends being gauged to access a top nozzle of a fuel assembly; anda tool head connected to the lower end of the tool body and sized to house a gripper assembly having a plurality of radially outwardly extending hooks in a withdrawn position so that the radially outwardly extending hooks are above and out of contact with the top nozzle of the fuel assembly when the tool head contacts or otherwise rests on the top nozzle, the gripper assembly being operable through an actuator to extend the plurality of radially outwardly extending hooks downward below its withdrawn position over a selected travel path that moves the hooks straight downward to an extended position where the hooks are pivoted outward during a last segment of downward extension to grip a portion of the top nozzle of the fuel assembly to support the fuel assembly as the crane or other hoist lifts the bail, with the gripper assembly being operable to reverse the selected travel path under the control of an operator when the fuel assembly is to be released, wherein the hooks for gripping the top nozzle of the fuel assembly have a laterally extending pin projecting from an upper end portion and the gripper assembly further includes:a carrier connected to the actuator, being raised and lowered by the actuator and pivotally connected to an upper end of the pin; anda cam slot on the tool head in which the laterally projecting pin on the upper end portion of the hook rides, the hook pivoting and raising or lowering as the carrier is raised or lowered by the actuator, wherein the pin riding in the cam slot first lowers the hook and then in a lower portion of the pin's travel in the cam slot the pin pivots the hook as the actuator is lowered. 2. The fuel assembly handling tool of claim 1 wherein the actuator is accessible from the bail plate and operable to extend or withdraw the gripper assembly to the extended or withdrawn position. 3. The fuel assembly handling tool of claim 2 wherein the gripper assembly is moved to the withdrawn position or the extended position by respectively raising or lowering the actuator in a linear motion. 4. The fuel assembly handling tool of claim 3 wherein the gripper assembly fully grips the fuel assembly as the actuator is lowered. 5. The fuel assembly handling tool of claim 2 wherein the gripper assembly positively locks in the withdrawn position and in the extended position. 6. The fuel assembly handling tool of claim 1 including guide pins extending down from the tool head for aligning the tool head with the fuel assembly top nozzle. |
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abstract | The present invention provides radioprotective materials and products produced using the materials, the materials and products being easily usable for blocking effects of radiations, exposure to which occurs unconsciously in daily life or in a working environment. Specifically stated, the present invention provides radioprotective materials containing polyamino acid(s) or derivative(s) thereof; the radioprotective materials in which the polyamino acid(s) comprises at least one amino acid selected from the group consisting of glutamic acid, phenylalanine, alanine and leucine; and the radioprotective materials in which the polyamino acid(s) or derivative(s) thereof is modified with sugar component(s). The present invention further provides radioprotective products produced using the radioprotective materials. |
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043326390 | claims | 1. In a nuclear reactor of the type having an internal core housed within a vessel, and a plurality of fuel assemblies housed within said core, each of said fuel assemblies including an open container having an inlet and outlet and an active substance such as plutonium oxide sealed within a relatively large number of elongated hollow pins located within said container, said reactor also including liquid metal heat exchanging fluid such as liquid sodium and means for circulating a stream of said fluid along a path, a section of which passes through said containers from their inlets to said outlets, a system for detecting breaks in said hollow pins, which breaks are of sufficient size to cause at least one predetermined contaminant to pass into said stream of fluid as the latter passes through said containers, said system comprising: first means including a first pump for collecting a combined sample of said fluid at the outlets of at least a group of said fuel assembly containers; first means for detecting the presence or absence of said contaminant in said combined sample, whereby to indicate the presence or absence of a break in at least one pin in said group of fuel assemblies; second means including a second pump having a lower flow rate capability than said first pump for selectively collecting individual samples of said fluid, one at a time, at the outlets of the fuel assembly containers in said group in the event said combined sample detecting means indicates the presence of a pin break said individual sample collecting second means including valve means for collecting said individual samples one at a time, said valve means including a main housing having wall means defining a plurality of spaced openings therethrough, said openings corresponding to and being in fluid communication with the outlets of said group of fuel assembly containers, respectively, a valve head located within said main housing and adapted for positioning in fluid communication with said openings individually, whereby to collect said individual fluid samples, and means for moving said valve head between said fluid communicating positions without engaging the inner surface of said wall means during movement of the valve head between said positions; and second means including a single detector for detecting the presence or absence of said contaminant in each of said individual samples, whereby to indicate the fuel assembly or assemblies responsible for the presence of said contaminant. a housing having openings corresponding in number to and in fluid communication with the outlet of at least a group of said fluid assembly containers for simultaneously receiving and combining heat exchanging fluid passing through the outlets of all of the containers in said group; and said assembly also including means for collecting a sample of said combined fluid for passing it to a first contaminant detection location and means for selectively collecting individual samples of said fluid, one at a time, from the outlets of said container group before the fluid is combined in said housing for passing said individual samples to a second contaminant detection location, said individual sample collecting means including a valve head located within said housing and adapted for positioning in fluid communication with said openings, individually, whereby to collect said individual samples, and means for moving said valve head between said fluid communicating positions without engaging the inner surface of said housing. providing a housing having openings corresponding to and in fluid communication with the outlets of at least a group of said fuel assembly containers, for simultaneously receiving and combining heat exchanging fluid passing through all of the containers in said group; collecting a sample of said combined fluid for passing it to a first contaminant detection location; and under predetermined circumstances, collecting individual samples of said fluid, one at a time, from the outlets of said container group before the fluid is combined in said housing and passing said individual samples to a second contaminant detection location, said individual sample collecting step including locating a valve head within said housing, positioning it in fluid communication with said openings, individually, whereby to collect said individual samples, and moving said valve head between said fluid communicating positions without engaging the inner surface of said housing. 2. A system according to claim 1 wherein said moving means includes first means for selectively moving said valve head to a position in confronting spaced relation with any of said openings, a predetermined distance from said inner surface, and second means for moving said valve head between any one of said confronting, spaced positions and a second position in engagement with the inner surface of said wall means surrounding the confronting, spaced positions and a second position in engagement with the inner surface of said wall means surrounding the confronting opening for placing said valve head in fluid communication therewith. 3. In a nuclear reactor of the type having an internal core housed within a vessel, and a plurality of fuel assemblies housed within said core, each of said fuel assemblies including an opened container having an inlet and outlet and an active substance such as plutonium oxide sealed within a relatively large number of elongated hollow pins located within said container, said reactor also including liquid metal heat exchanging fluid such as liquid sodium and means for circulating a stream of said fluid along a path, a section of which passes through said containers from their inlets to their outlets, a valve assembly for use in a system for detecting breaks in said hollow pins, said assembly comprising: 4. An assembly according to claim 3 wherein said moving means includes first means for selectively moving said valve head to a position in confronting spaced relation with any of said openings, a predetermined distance from the inner surface of said housing, and second means for moving said valve head between any one of said confronting, spaced positions and a second position in engagement with the inner surface of said housing surrounding a confronting opening for fluid communication therewith. 5. In a nuclear reactor of the type having an internal core housed within a vessel, and a plurality of fuel assemblies housed within said core, each of said fuel assemblies including an opened container having an inlet and outlet and an active substance such as plutonium oxide sealed within a relatively large number of elongated hollow pins located within said container, said reactor also including liquid metal heat exchanging fluid such as liquid sodium and means for circulating a stream of said fluid along a path, a section of which passes through said containers from their inlets to said outlets, a method of collecting fluid samples for use in a system for detecting breaks in said hollow pins, said method comprising: 6. A method according to claim 5 wherein said moving step includes selectively moving said valve head to a position in confronting spaced relation with any of said openings, a predetermined distance from the inner surface of said housing and moving said valve head between any one of said confronting, spaced positions and a second position in engagement with the inner surface of said housing surrounding of the confronting opening for fluid communication therewith. |
description | This application is a continuation of U.S. patent application Ser. No. 14/129,504 filed Apr. 4, 2014, now U.S. Pat. No. 9,190,181 which was published on Aug. 7, 2014 under Publication No. US2014/0221722 A1, which is United States National Phase of PCT Patent Application No. US2012/045084 filed on Jun. 29, 2012, which was published on Jan. 3, 2013 under Publication No. WO 2013/003796 A2, which claims priority to U.S. Provisional Patent Application No. 61/502,557 filed Jun. 29, 2011, which are incorporated herein by reference. Nuclear reactors generate 19 percent of the electricity in the U.S., and this process generates high-level radioactive waste in the form of uranium oxide or mixed oxide fuels. Approximately 1000 m3 (6200 bbl) of high-level waste is produced each year from commercial reactors in the U.S., and additional material is generated by military operations. Europe is also heavily invested in nuclear power (e.g. more than three-fourths of the electricity in France is generated by nuclear reactors), and other countries worldwide have started to aggressively pursue nuclear energy to power their growing economies. As a result, the current rate of nuclear waste generation is approximately 10,000 m3/yr, and the amount of radioactive waste being generated worldwide is expected to increase significantly. Yet, there are no safe, reliable ways to dispose of nuclear waste on site, that is, at the source of the waste's generation. This waste includes but is not limited to spent nuclear fuel from nuclear reactors, high-level waste from the reprocessing of spent nuclear fuel, transuranic waste mainly from defense programs, and uranium mill tailings from the mining and milling of uranium ore. High-level nuclear waste is currently stored at the reactor where it was generated. The only serious options for disposal being considered are to place the waste in low permeability geologic formations, like tight rock or clay. The current approach for disposal of radioactive waste is not without problems. Congress has mandated a 10,000-year period of isolation, but it is difficult to guarantee that waste at the shallow depths of current repositories will remain isolated from the biosphere, or human intervention, for even a fraction of this time. Yucca Mountain, a 300-m-deep facility near Las Vegas, is the only U.S. option for high-level waste disposal. This facility has been scrutinized for 20 years, and even after a $50B expenditure the earliest it could open is 2017. Considerable political opposition by Congress, the state of Nevada and others may delay opening even further. For example, Congress did not provide any funding for development of the site in the 2011 federal budget. Significant uncertainty exists about the feasibility of waste placed at a depth of 300 m remaining isolated from the biosphere for 10,000 years, and this uncertainty is the basis for much of the opposition to Yucca Mountain. Even if Yucca Mountain does open, all its capacity has been allocated and options for additional capacity are being considered. The politics involved in finding permanent disposal sites is, at best, difficult and, at worst, intractable. Because the waste remains radioactive for a very long time, no one wants this waste traveling through their “backyard” on its way to a permanent disposal site or in their “backyard” as the disposal site. As politicians and the public continue to debate the issue, the waste remains temporarily stored on site in ways that are arguably far less safe than any proposed permanent disposal solution. For example, nuclear reactors temporarily store the waste on site in water pools. The devastating earthquake and tsunami in northeast Japan, which knocked out power sources and cooling systems at Tokyo Electric Power Co's Fukushima Daiichi plant, demonstrates how tenuous and potentially dangerous this storage practice really is. Therefore, a need exits for a safe, reliable method of disposing nuclear waste on site and one that could achieve the 10,000 year period of isolation required by Congress and sought by other countries. A system and method according to this invention involves storing nuclear waste or hazardous waste in hydraulic fractures driven by gravity, a process referred to herein as “gravity fracturing.” For the purposes of this disclosure, nuclear or radioactive waste is considered a hazardous waste although in the environmental industry radioactive waste is often not labeled as “hazardous waste.” The method creates a dense fluid containing waste, introduces the dense fluid into a fracture, and extends the fracture downward until it becomes long enough to propagate independently. The fracture will continue to propagate downward to great depth, permanently isolating the waste. Storing solid wastes by mixing the wastes with fluids and injecting them into hydraulic fractures is a well-known technology in the petroleum industry. Nuclear waste was injected into hydraulic fractures at Oak Ridge in the 1960s. The essence of the invention differs from conventional hydraulic fracturing techniques in that it uses fracturing fluid heavier than the surrounding rock. This difference is fundamental because it allows hydraulic fractures to propagate downward (rather than horizontally) and carry wastes by gravity instead of by pumping. More specifically, the method of disposing nuclear waste and other hazardous waste includes the steps of blending the waste with water or other fluid and a weighting material to make a dense fluid or slurry of a predetermined density, temperature and viscosity; and injecting the dense fluid or slurry—at a predetermined pressure and/or rate into a well so that the fluid or slurry enters the strata at a predetermined depth and continues to travel downward through the strata until the fluid or slurry, becomes immobilized. Prior to the blending step, the waste, if in solid form, may be ground into particles of a predetermined size. The pressurized blended mixture cracks and dilates the rock structure, which is preferably a stable, low permeability rock structure such as many igneous and metamorphic rocks as well as some sedimentary rocks. (Initially, propping the fracture is avoided). Because the dense fluid has a density greater than that of the rock, the fluid or slurry has an absolute tendency to travel downward by gravity (until the density relationship changes or other mechanics arrest the downward travel) and remain far below the earth's surface. The dense fluid may include water, oil, gel or any fluid suitable for providing the required viscosity and density. The well is preferably drilled at and on the site which generates the nuclear waste or other hazardous waste, thereby eliminating the need to transport the waste off-site and to the disposal site. The well includes a work string or tubing for receiving the blended fluid, waste and weighting material; a packer; and a cemented steel casing with perforations located at or about the predetermined depth. The predetermined depth is preferably in a range of about 10,000 to 30,000 feet (about 3,000 to 9,000 meters) but it can be shallower or deeper depending upon rock properties and drilling limitations. The weighting material may be other nuclear waste (including, for example, radionuclides such as uranium), other hazardous waste or a metal such as bismuth, lead, or iron in order to add weight to the primary waste which is being disposed. Metals or alloys that are in liquid phase at the temperature and pressure encountered in the subsurface are particularly suitable as a weighting material. The work string may be pulled for routine cleaning or replacement. The blender used to blend the water, waste and weighting material is preferably shielded, as is the pumping unit (e.g., a pumping truck) used to pump the mixture at pressure into the well. Although not illustrated, the dense fluid may propagate downward and then curve in a horizontal direction creating a sub-horizontal storage space. Hydraulic fractures are created when the pressure in a fluid-filled crack causes the material at the crack tip to fail. The fracture advances and fluid flows forward to fill the newly created space. Hydraulic fractures are commonly created by using a pump to inject fluid into a well, but this is by no means the only occurrence. Geologic examples are well known in which hydraulic fractures grow upward through the Earth's crust because the fractures are filled with liquid lighter than their enveloping rock. A dike filled with magma that propagates upward to feed a volcanic eruption is one example of a hydraulic fracture propagating by gravity. A system and method according to this invention involves propagating hydraulic fractures downward by filling the fractures with dense fluid containing waste. Propagation occurs when the pressure in the fracture creates a stress intensity that exceeds the toughness or strength of the rock. Referring to FIGS. 1 to 3, an open borehole is created and filled with the dense fluid until the pressure at the bottom is sufficient to create a fracture (FIG. 3 at “a”). A similar fracturing process occurs during overbalanced drilling when the mud weight is too great and causes circulation to be lost by initiating a fracture and causing it to grow away from the borehole. Fluid will flow into the fracture and the level of fluid in the well will drop (FIG. 3 at “b”). However, the fracture is expected to advance faster than the rate of drop of fluid level in the well, so the overall height from the tip of the fracture to the top of the fluid column in the well lengthens. This increases the driving pressure and furthers downward propagation as the fluid in the wellbore drains by gravity into the fracture (FIG. 3 at “c”). The vertical span of the fracture continuously increases, causing the pressure at the bottom of the fracture to increase and ensuring continued downward propagation, even after all the liquid has drained from the well into the fracture (FIG. 3 at “d”). The pressure distribution causes the lower part of the fracture to bulge open and the upper part to pinch shut. A residual coating of fluid will be left behind when the fracture closes, and this will diminish the volume of fluid in the fracture. Eventually the original fluid will be spread as a thin coating on the fracture wall, extending from the bottom of the borehole to great depth. In the case of slurry, the fracture may be propped if the liquid leaks off into the rock. The process is repeated by putting additional fluid into the well. This will create a new fracture that will follow the path of the earlier one (FIG. 3 at “e”). The additional fluid reaches an even greater depth than the original batch. The maximum depth that can be reached by dense fluids is unclear, but it could exceed tens of kilometers. A method of disposing nuclear waste and other hazardous waste practiced according to this invention, therefore, effectively removes the waste from exposure to human activities at a time scale relevant to both societial actions and the half-lives of many hazardous radionuclides. The method includes the steps of blending the waste with materials suitable for creating a dense fluid or slurry which has a predetermined density and viscosity; and injecting the dense fluid at a predetermined pressure or rate into a well so that the dense fluid enters the strata at a predetermined depth and continues to travel downward through the strata until its flow stops, for example, because the solid-to-liquid ratio is too high to allow flow. Propagation may also stop when a sufficient amount of the dense fluid or fluid/slurry has been spread as a film or residue over the upper closed portion of the fracture. Oil, gel or any fluid suitable for providing the required viscosity and density may be used Weighting material adds density to the primary waste which may be other types of nuclear waste, other hazardous waste or a metal such as, but not limited to, bismuth, lead, iron, copper, or low melting point metals or alloys (e.g., mercury, woods metal, indalloy 15, gallium) that could mix with and possibly dissolve or amalgamate high-level waste material. The low-melting-point alloys are a liquid under the expected pressure and temperature conditions at the bottom of the injection well. Solid compounds such as metals used for weighting material may be mixed with a high-shear-strength liquid, including polymer gels that may be crosslinked, or inorganic gels that may formed by hydrating clay minerals, to create a dense slurry. Prior to the blending step, the waste, if in solid form, may be ground to a predetermined size. The pressurized dense fluid creates a vertical fracture or crack in the rock structure. The dense fluid enters the crack and serves to prop the rock structure. The rock structure is preferably a stable, low permeability rock formation, of the kind that nuclear reactors are typically built over and upon. Because of the weighting material, the density of the dense fluid is greater than that of the rock and this causes an absolute tendency for the fluid to travel downward until it becomes immobilized. If the density of the dense fluid is exactly equal to that of the rock, the dense fluid may be unable to overcome the rock fracture toughness. This is required for fracture propagation, hence the density should be somewhat higher to ensure the fracture growth. How much higher depends upon the fracture toughness magnitude, fluid properties, and other effects standard in industrial hydraulic fracturing. In general terms, the density of rock increases as depth increases. Therefore, once the fracture propagates, a point can be reached where the density of the dense fluid becomes the same as the density of the rock, thereby limiting any further propagation downward. Eventually, the fracture becomes sub-horizontal and the dense fluid fills the fracture horizontally. This is similar to geological sills and does not hamper the proposed technology as the horizontal part of the growing fracture also allows for safe waste storage. Fracture toughness also increases with depth because it increases with such factors as temperature, pressure and size of the fracture. However, the effect of fracture toughness can be overcome by pressurizing the fracture. For example, and just by way of example the immobilization point may occur at about 2,000 to 50,000 feet (about 600 to 15,000 meters) below the dense fluid's initial entry point into the strata. (The depth can be greater and is mostly constrained by drilling and pumping limitations.) The dense fluid can be monitored by using conventional tracer means to see whether any movement or migration has occurred upward relative to the perforations in the well casing, or it can be monitored using microseismics means to evaluate downward migration below the bottom of the region accessible to the well casing. The well is preferably drilled at and on the site which generates the nuclear waste or other hazardous waste, thereby eliminating the need to transport the waste off-site and to the disposal site. The well also eliminates the need for temporary storage means on site because the waste can be transported directly to the well for immediate permanent disposal. As shown in FIG. 2, the well includes a work string or tubing for receiving the blended water, waste and weighting material; a packer; and a cement casing with perforations located at or about the predetermined depth. The predetermined depth is preferably in a range of about 10,000 to 30,000 feet (about 3,000 to 9,000 meters). The work string may be pulled for routine cleaning or replacement. The blender used to blend the water, waste and weighting material is preferably shielded, as is the pump truck used to pump the dense fluid at pressure into the well (see FIG. 1). Preferred embodiments of a system and method for abyssal sequestration of nuclear waste and other types of hazardous waste have been described and illustrated, but not all possible embodiments. The inventive system and method itself is defined and limited by the following claims. |
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abstract | A flexure carriage assembly has a carriage formed of a substantially rigid material. The carriage has four elongate columns arranged spaced apart and parallel to one another. Each of the elongate columns has first and second ends. The flexure carriage has four first cross members disposed between adjacent pairs of elongate columns and arranged to interconnect the first ends. The flexure carriage also includes four second cross members arranged between adjacent pairs of elongate columns and arranged to interconnect the bottom ends. The elongate columns and first and second cross members define a three-dimensional rectangular structure. The flexure carriage also has disposed centrally between the four elongate columns a translating section spaced equidistant between the first and second ends of the columns. A plurality of flexures are disposed between the translating element and elongate columns and between the elongate columns and first and second cross members in order to permit precise movement of the translating section in a plane according to applied forces against edges of the translating section. A pair of piezoelectric assemblies are connected to the translating section. One applies force to the translating section in a first linear path and the other applies force to the translating section in a second linear path perpendicular path. |
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060027347 | summary | FIELD OF THE INVENTION The invention relates generally to assaying ore samples for valuable components (assay elements), and more specifically to a technique of assaying large ore samples for gold, silver, and barium content using photon activation analysis. BACKGROUND OF THE INVENTION Classical fire assaying is and has remained the standard assay for gold and silver in precious metal mining. In this technique, a small rock sample, typically 30-150 grams (gm) is mixed with litharge (lead oxide) and silica fluxes and fused in a high temperature furnace. A resultant lead button containing solubilized gold and silver is poured into a conical mold. This lead button is subsequently cupeled in a bone ash or other cupel whereby the molten lead is converted to lead oxide which is absorbed by the substance of the cupel. When the cupeling is completed, a small dore bead of gold and silver remains. This bead is parted in nitric acid to remove the silver after recording the initial weight. The parted bead is weighed again giving the gold weight directly. The silver weight is found by difference of the dore bead weight and the gold bead weight. Alternatively, the dore bead can be dissolved in aqua regia and the resultant solution analyzed on an atomic absorption (A.A.) instrument. The original sample may also be digested in aqua regia to solubilize the gold and silver, and the resultant solution analyzed on an A.A. instrument. In this technique a small sample, typically 10 to 15 grams (gm), is digested in the aqua regia. Some difficulty is encountered with samples that contain very low quantities of gold and silver. Several samples may be fused and the resultant beads added together to obtain greater sensitivity, but this required more time and extra cost. In addition, very great care is required in temperature control and timing to obtain accurate silver assays. While fire assaying is a very sensitive assay technique for gold particularly, which is why it has been used for such a long time, its sensitivity is limited by the size of bead that can be detected and weighed or solubilized in an A.A. finish. There are some organic solvents, such as M.T.B.K., which can extract gold from an aqueous solution and thus concentrate the gold for greater sensitivity on an A.A. instrument. Additionally, all of these techniques suffer from allowing only small ore samples to be analyzed. In order for the sample data to be meaningful, the small sample must be representative of the larger and typically non homogeneous rock sample from which it is obtained. To date, the only technique which has allowed this representativeness has been a meticulous sample preparation wherein a large rock sample is crushed into successively smaller sizes before being split into a smaller sample. The final sample is pulverized typically to less than 150 mesh prior to be digested or fused for assay. This meticulous sample preparation is not only expensive, but it can be very difficult to achieve with native gold metallics present. Gold is usually not homogeneously distributed in a sample but occurs most commonly as metallics frequently alloyed with silver. While silver can occur as a native metallic, it is more frequently present with gold as electrum or combined with sulfur as a silver sulfide. The problems are aggravated when the ore deposits are of low grade. The current price of gold and the development of process and mining technology has allowed the development of large low grade deposits. These deposits typically require much drilling and blasting to break the ore for processing. Large volumes of samples are prepared from these blast holes for assaying to control the mining. The fire assay technique has remained the standard, and low throughput, low sensitivity for low gold content, and the requirement for meticulous sample preparation have been accepted as inevitable. In view of the above discussion, and object of the present invention is to provide a system which can assay relatively large samples of ore for gold and other assay elements of interest. An additional object of the invention is to provide an assay system for certain elements in various environments which requires minimal sample preparation. Yet another object of the invention is to provide an assay system with throughput which is greater than the prior art fire assay method. Another object of the invention is to provide an assay system with sensitivity sufficient to meet requirements for commercial ore testing and production standards. Still another object of the invention is to provide an assay system whose accuracy and precision is not adversely affected by typically non homogeneous ore samples and especially relatively large, non homogeneous samples. Another object of the invention is to provide an assay system that is relatively insensitive to sample geometry. Yet another object of the invention is to provide a non destructive assay or analysis system that can be used with a wide variety of sample types to determine the elemental concentration of any element susceptible to detection by means of photon activation. There are other objects and applications of the invention which will become apparent in the following disclosure. SUMMARY OF THE INVENTION The present invention provides a high throughput technique that permits large sampled of materials to be analyzed for their gold, silver, and other assay element contents, thereby avoiding the time and cost necessary for a meticulous sample preparation. The technique is effective even when the distribution of metallics in the sample is not homogeneous. The present invention departs from the prior art techniques in that it recognizes that certain nuclear physics techniques can yield assay information. Specifically, the invention exploits the properties of certain elements having isotopes with one or more excited nuclear states that are characterized by relatively long half lives (microseconds to minutes). Nuclei in these excited states are referred to as isomers, and cannot be produced directly from the ground state. Rather they must be produced by exciting the nucleus to a higher excited state which quickly decays to the long-lived isomeric state. The isomers decay to the ground state through the emission of a gamma ray having a well defined energy for the particular element. Typical gamma ray energies for isomeric transitions are in the range of 0.05-1.0 million electron volts (MeV) or 50-1000 thousand electron volts (KeV). The technique of the invention entails irradiating the sample with a beam or flux of gamma rays of sufficient energy to excite the nuclei of the assay elements into their isomeric states, ceasing the irradiation, detecting and identifying the gamma rays resulting from the decay of the isomeric states to the ground state, and the analyzing the detected gamma rays to determine the content of assay elements in the sample. In a preferred embodiment, the irradiated sample is rapidly moved to a shielded low background environment in which the gamma rays from the isomeric transition are detected. As mentioned previously, the ore samples are typically non homogeneous. As an example, gold ore can contain gold in highly concentrated "nuggets" which are rather sparsely and non homogeneously distributed through a large volume of non gold bearing material. The average gold concentration of such an ore may, however, be well above the commercial threshold. Any meaningful assay system must be able to accurately obtain results for non homogeneous ore. To address this problem, the present invention is embodied so that portions or "segments" of the ore sample are sequentially irradiated and counted. During each irradiation, the sample is oscillated within the gamma ray flux in order to obtain uniform exposure of each portion. After all segments of the sample have been irradiated and counted, the count results are combined in order to obtain a highly representative assay of the entire sample volume. The technique has a similar sensitivity to gold as fire assay, 0.001 troy ounce per short ton, but can quickly handle large sample weights (typically 10 kg) to give better average assay numbers. The technique makes it possible to process large volumes in a reasonable time (more than 600 samples within 24 hours). While the nuclear physics phenomena exploited by the invention are known, the present discussion of the physics underlying the invention technique is not intended as an admission that the nuclear physics phenomena were recognized in the prior art as having any applicability to a technique for assaying large ore samples, or having applicability to any other stated objects of the invention. The system can also be configured to analyze many types of samples non destructively for any element which is subject to photon activation analysis. |
abstract | An irradiation treatment device is provided in which uniform irradiation treatment with high speed can be achieved without damaging a substrate, even in the case in which the substrate to be treated is enlarged and exceeds the length of the rod-shaped dielectric barrier discharge lamp. The object is achieved in a substrate treatment device using dielectric barrier discharge lamps in which the dielectric barrier discharge lamps and the substrate are transported relative to one another and in which the surface of this substrate is irradiated with UV light from the dielectric barrier discharge lamps, in that the length for the above described dielectric barrier discharge lamps in the lengthwise direction is less than the length in the direction perpendicular to the transport direction of the substrate. There are at least two dielectric barrier discharge lamps and there is an area of the above described substrate which has been irradiated by one dielectric barrier discharge lamp and there is an area of the above described substrate which has been irradiated by the other dielectric barrier discharge lamp during transport of this substrate such that they come to rest on one another at least in one part, and that with respect to the UV light emitted by the respective dielectric barrier discharge lamp in this area in which superposition occurs, there are light screening means by which a transition is effected between the two lamps. |
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051065703 | claims | 1. A method for generating a negative hydrogen ion beam comprising the steps of: (a) providing a source of metal hydride; (b) heating said source to dissociate said metal hydride and thereby generating one of a quantity of hydrogen atoms and a quantity of negative hydrogen ions at a surface of said hydride; (c) applying a negative electric charge to said surface of said metal hydride to ionize said hydrogen atoms and to maintain said ions as negative hydrogen ions; and (d) electrically accelerating said hydrogen ions as a directed beam. (a) providing a source of metal hydride; (b) directing an electron beam against a surface of said source and thereby generating negative hydrogen ions at said surface; and (c) electrically accelerating sad hydrogen ions as a directed beam. 2. The method of claim 1 wherein said hydrogen ions are electrically accelerated by the application of a d.c. electric field at said surface of said metal hydride. 3. The method of claim 1 wherein said step of applying an electrical charge includes directing an electron beam onto said surface of said hydride. 4. The method of claim 1 wherein said metal hydride is selected from the group consisting of the AB.sub.5 hydrides, binary metal hydrides, the saline hydrides, and the alkaline earth metal hydrides. 5. A method for generating a negative hydrogen ion beam comprising the steps of: 6. The method of claim 5 further comprising the step of applying a negative electric charge to said surface of said metal hydride to prevent said negative hydrogen ions from losing charge. 7. The method of claim 5 wherein said hydrogen ions are electrically accelerated by the application of a d.c. electric field at said surface of said metal hydride. 8. The method of claim 5 wherein said metal hydride is selected from the group consisting of the AB.sub.5 hydrides, binary metal hydrides, the saline hydrides, and the alkaline earth metal hydrides. |
abstract | A travelling wave reactor for a space exploration. A reactor core of the travelling wave reactor is dispersed into several modules in a travelling wave direction; a new reactor is sequentially provided with a starting source module and a plurality of new fuel modules at zero burnup; all the modules are coaxially assembled in the travelling wave direction by means of an assembling parts, and each module further includes a heat pipe; the heat pipe in each module positioned at a front part sequentially passes through all the modules positioned at a rear portion thereof and extends out of the module at a rear end; and after a period of time of burn-up, the reactor core of the travelling wave reactor is provided with the starting source module, a spent fuel module, a critical fuel module and the new fuel module sequentially in the travelling wave direction. |
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047042354 | description | The invention will now be described in conjunction with some non-limiting examples. EXAMPLES In a number of decontamination tests carried out on samples of Inconel 600 taken from a number of different positions of steam generator in a PWR after an operation time of about 8 years, which samples were the most difficultly decontaminable ones to be obtained, the samples were treated at room temperature (about 20.degree. C.) for 48 hours in an oxidizing solution in accordance with the invention. This solution consisted of an aqueous solution, made acidic to pH 1.4 with nitric acid, of 1.5 g/l of cerium(3)nitrate, 0.1 g/l of chromic acid and 12 g/l of boric acid to which ozone was continuously supplied. The decontamination factors obtained at these experiments were 20-300. In comparative tests carried out on similar samples, i.e. tubes of Inconel 600, with the most effective of the soft processes tested at the Agesta and Surry-II projects, both requiring an operational temperature of 80.degree.-90.degree. C., decontamination factors with an average of merely 6.2 were obtained. In addition to the decontamination tests reported above, corrosion tests have been carried out on blank, non-preoxidized samples of Inconel 600. The test pieces were 3 pieces of steam generator tubes with lengths of 5 cm. To simulate the condition in the rolled zone in the tubes of the tube plate these samples had been rolled internally to half the lengths thereof. The cold deformation obtained was about 5%. The three samples were exposed in parallel in a 100 ml glass container in an aqueous solution containing 12 g/l of boric acid, 1.5 g/l of Ce(3)nitrate and 0.1 g/l of CrO.sub.3 in nitric acid at a pH of about 1.4. Oxygen gas with about 2.5 percent by volume of ozone was bubbled into the same container at a rate of about 0.1 l/min. The temperature was 20.degree. C. and the exposure time was 48 hours. The weight losses during this exposure are accounted for in the Table below. A further exposure cycle, identical with the first one was carried out. The weight losses at this exposure are also presented in the Table. The material loss after each exposure cycle is on an average well below 1 .mu.m, which must be regarded as extremely satisfactory. No signs of local corrosion have been observed. TABLE ______________________________________ Corrosion tests carried out on 3 identical rolled tube samples of Inconel 600 exposed in parallel to each other. Exposure times Cycle Time hours ______________________________________ 1a 48 1b 48 ______________________________________ Tube sample Material loss (.mu.m) No. Cycle 1 Cycle 2 ______________________________________ 0 0.86 0.68 1 0.91 0.58 2 0.98 1.11 ______________________________________ |
abstract | A positioning stage system for precise and accurate movement of an article in an electron beam lithography system. The positioning stage system includes a support platform for supporting the article, an X-direction linear motor coupled to an X-member, a Y-direction linear motor coupled a Y-member, and a slide attached to the support platform and slidably engaged to the X-member and the Y-member. The X-member and Y-member, upon actuation of the X-direction and Y-direction linear motors, cause the support platform to move in an X-direction and a Y-direction, respectively. |
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abstract | A system for use in shutting down a nuclear reactor includes a housing that defines a region therein sealed from an ambient environment and a gate member disposed within the region in a manner such that the gate member segregates the region into a first compartment and a second compartment isolated from the first compartment. The gate member is formed from a material having a predetermined melting point. The system further includes a neutron absorbing material disposed within the first compartment and a dispersion mechanism disposed within the region. The dispersion mechanism structured to encourage the neutron absorbing material from the first compartment into the second compartment. |
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abstract | The present invention provides a corrosion-resistant structure for a high-temperature water system comprising: a structural material 1; and a corrosion-resistant film 3 formed from a substance containing at least one of La and Y deposited on a surface in a side that comes in contact with a cooling water 4, of the structural material 1 which constitutes the high-temperature water system that passes a cooling water 4 of high temperature therein. Due to above construction, there can be provided the corrosion-resistant structure and a corrosion-preventing method capable of operating a plant without conducting a water chemistry control of cooling water by injecting chemicals. |
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description | The present invention relates to a method for checking for leakage from tubular batteries in which occurrence of leakage from the sealed end face of a tubular battery is checked by X-ray fluorescence analysis after its manufacture. Tubular batteries such as cylindrical or prismatic batteries are sealed, e.g., by inwardly crimping the open end portion of a bottomed tubular battery case to compress an insulating gasket, thereby providing liquid-tight sealing between the battery case, the insulating gasket, and a sealing member. However, even a trace amount of electrolyte deposited on the sealed portion or an imperfect sealing itself would cause the interface between two components in the sealing structure to be wetted with the electrolyte. This may lead to the development of a leakage path, resulting in leakage occurring by the electrolyte migrating along the leakage path. In particular, in those batteries with an alkaline electrolyte, the alkaline electrolyte itself migrates along the surface of the negatively charged metallic sealing member or battery case. Leakage is thus more likely to happen in these batteries when compared with other types of batteries. Conventionally, to check for the occurrence of leakage, a given number of tubular batteries were arranged side by side with their sealed end faces oriented upwardly. The sealed end face of each tubular battery was then covered with a cloth, to which a reagent was then applied. While the cloth was being tapped with a brush, the tubular battery whose yellow reagent discolored purple was visually identified and determined to be leaky. However, such inspection means that relies on operators' manual operations and visual determinations is limited in terms of the handling speed and thus very inefficient. The inspection is also likely to be inaccurate because variations exist among individual operators and they may overlook leakage. To be worse, such a leakage that occurs immediately inside the sealed end face cannot be visually identified. In this context, such inspection means has been recently adopted which allows for determining occurrence of leakage by X-ray fluorescence analysis. According to a first one of those conventional techniques, a primary X-ray having a certain pre-defined wavelength is used to irradiate a tubular battery, and a fluorescent X-ray coming out of the tubular battery is allowed to be incident upon an analyzer. The analyzer analyzes whether the incident fluorescent X-ray contains such a fluorescent X-ray that has a wavelength associated with an electrolyte component. Then, based on the output from the analyzer, occurrence of leakage is determined (for example, see Patent Document 1). On the other hand, according to a second conventional technique, while tubular batteries are being fed in a row at predetermined intervals, each tubular battery is irradiated with an X-ray from an X-ray source. Those fluorescent X-rays arising from the sealed end face and a side of a tubular battery are allowed to be incident upon a plurality of detectors located around the X-ray source. Based on the results of detections provided by the plurality of detectors, a tubular battery with the electrolyte deposited thereon is identified (for example, see Patent Document 2). [Patent Document 1] Japanese Patent Laid-Open Publication No. Sho 52-138627 [Patent Document 2] Japanese Patent Laid-open Publication No. Hei 9-203714 However, any of the aforementioned conventional techniques cannot check tubular batteries at high speeds with accuracy. That is, the first conventional technique cannot increase the speed of inspection because batteries are checked one by one with a battery placed opposite to the X-ray source. On the other hand, in the second conventional technique, four detectors placed in a rectangular arrangement are used also to detect the electrolyte deposited on a side of the tubular batteries, thereby requiring the batteries to be arranged at relatively large intervals. This arrangement places a limitation on the speed of inspection as well as causes an increase in the size of the system and costs. Furthermore, in the inspection of leakage from a battery that employs an alkaline electrolyte, it is commonly practiced to analyze the intensity of a fluorescent X-ray emitted from the potassium in the alkaline electrolyte. However, with the first and second conventional techniques, the fluorescent X-ray is incident upon the detector through a path in which air is present. This raises a problem that an element that is contained in the air, especially argon, emits-a fluorescent X-ray of a wavelength similar to that of potassium, thereby exerting an adverse effect on the detection of the intensity of the potassium to degrade the accuracy of detection. The present invention was developed in light of the aforementioned conventional problems. It is therefore an object of the present invention to provide a method for checking for leakage from tubular batteries, in which occurrence of leakage from tubular batteries is determined at high speeds with accuracy by X-ray fluorescence analysis. To achieve the aforementioned object, a method for checking for leakage from tubular batteries according to the present invention includes: feeding tubular batteries with respective axial centers thereof aligned in parallel to each other to pass through a leakage check section placed opposite to a detection window of a leakage check mechanism; irradiating a sealed end face of the tubular battery in the leakage check section with an X-ray through the detection window and allowing a fluorescent X-ray coming out of the sealed end face to enter a fluorescent X-ray detector through the detection window; and analyzing whether a fluorescent X-ray associated with an electrolyte component is contained in the incident fluorescent X-ray to thereby determine whether leakage occurs from the tubular battery. In the method, the detection window is defined in such a shape that a length thereof in a direction of feed of the tubular batteries is less than a spacing between the tubular batteries being fed, and a length thereof in an orientation orthogonal to the direction of feed is slightly larger than an outer size of the cross-sectional shape of the tubular batteries in an orientation orthogonal to their axial center. In such an arrangement, when the tubular batteries with their respective axial centers aligned in parallel to each other pass through the leakage check section opposite to the detection window of the leakage check mechanism, a fluorescent X-ray emitted from the sealed end face of the tubular batteries irradiated with the primary X-ray enters the fluorescent X-ray detector through the detection window. Since the detection window is defined in such a shape that its length in the direction of feed of the tubular batteries is less than the spacing between the tubular batteries, the fluorescent X-ray coming out of each of two adjacent tubular batteries will never simultaneously enter the fluorescent X-ray detector through the detection window. Accordingly, even when the tubular batteries arranged at the smallest possible spacing are each fed at a high speed, it is ensured that the fluorescent X-rays emitted from individual tubular batteries and incident upon the fluorescent X-ray detector are separately identified, thereby allowing for dramatically increasing the speed of checking for leakage. On the other hand, the length of the detection window in an orientation orthogonal to the direction of feed is defined to be slightly greater than the outer size of the cross-sectional shape of the tubular batteries in an orientation orthogonal to their axial center. This ensures that the fluorescent X-ray emitted from any part of the sealed end face of the tubular batteries enters the fluorescent X-ray detector, thereby making it possible to detect occurrence of leakage wherever the leakage has occurred on the sealed end face. It is also possible to detect a predetermined fluorescent X-ray incident upon the fluorescent X-ray detector with high accuracy at a high S/N ratio. Furthermore, since the X-ray and fluorescent X-ray pass through the insulating gasket or the like, it is possible to positively detect even such a leakage occurring inside the tubular batteries that could not be visually determined. Furthermore, occurrence of leakage may be detected in accordance with an intensity per unit time of the fluorescent X-ray successively entering the fluorescent X-ray detector from each tubular battery which sequentially comes to oppose the detection window while being fed, or in accordance with an intensity per unit area of the sealed end face of the tubular battery. This makes it possible to highly accurately detect occurrence of leakage from the tubular batteries being fed even at high speeds no matter how each of the tubular batteries arranged in position is fed, i.e., continuously at a constant speed, intermittently at a standstill in the leakage check section where the tubular battery opposes the detection window, or variably at a low speed only when the tubular battery passes through the leakage check section. This is because occurrence of leakage is determined in accordance with either one of the following intensities. That is, the intensities include the strength or a fluorescent X-ray per unit time that is determined by dividing the fluorescent X-ray incident from a tubular battery when passing through the leakage check section by the time required for the entire tubular battery to completely pass by the detection window. The intensities also include the strength of a fluorescent X-ray per unit area that is determined by dividing the fluorescent X-ray incident from a tubular battery when passing through the leakage check section by the surface area of the sealed end face of the tubular battery. Furthermore, the detection window of the leakage check mechanism may be disposed to oppose the sealed end face of the tubular battery being fed at a predetermined distance therebetween. The housing of the check mechanism may contain an X-ray source for emitting an X-ray to a tubular battery, a mask for condensing a fluorescent X-ray emitted from the X-ray source into a beam, and the fluorescent X-ray detector upon which the fluorescent X-ray is incident. Additionally, the inside of the housing may be kept in a helium gas atmosphere, thereby allowing the helium gas filled in the housing to reduce the argon gas contained in the air. It is thus possible to remove adverse effects exerted by the argon gas on the fluorescent X-ray, thereby eliminating noise caused by the argon gas and highly accurately detecting the strength of the fluorescent X-ray at a high S/N ratio. However, in this case, to prevent leakage of the helium gas out of the detection window, it is preferable to close the detection window with a sealing member made of a material that transmits the X-ray. Furthermore, since the mask can condense the primary X-ray into a beam, the opening area of the detection window can be reduced as small as possible. Additionally, the distance between the detection window and the sealed end face of the tubular batteries being fed through the leakage check section can be set, for example, to be as small as approximately 2 mm. It is thus possible to reduce the adverse effects caused by an argon gas contained in the air that is present between the detection window and the sealed end face of the tubular batteries, thereby improving the accuracy of detecting leakage. Furthermore, the tubular batteries may be fed while being held on transfer disks in parallel to each other at regular intervals. This allows each of the tubular batteries arranged and held on the transfer disks in the predetermined manner to be fed in a rotary scheme toward the leakage check section. Accordingly, unlike a case where the tubular batteries disposed upright are fed on a conveyor, there is no possibility of the tubular batteries toppling over, and thus the speed of feed can be significantly increased. Furthermore, even when the tubular batteries are each fed at high speeds, the transfer disks positively hold the tubular batteries so that they are not displaced out of position while being fed at the high speeds. This ensures that the sealed end face of the tubular batteries can pass by the detection window at the shortest possible constant distance therebetween. It is thus possible to detect leakage from the tubular batteries with high accuracy while they are beings fed at high speeds. Furthermore, the housing which accommodates the X-ray source, the mask, and the fluorescent X-ray detector may be installed in front of an apparatus casing, to which the transfer disks for feeding the tubular batteries are attached, so that the detection window provided on the housing opposes the transfer disks. In this arrangement, the housing is attached to the support mount so that the detection window which transmits the primary X-ray is oriented backwardly in consideration of the operator normally working only in front of the support mount opposite to the attachment of the transfer disks. This arrangement allows for eliminating the risk of the operator being exposed to the primary X-ray, thereby embodying a very safe leakage check system in practically realizing a leakage check method of the present invention. Furthermore, each of the tubular batteries may be held in position on the transfer disks to pass through the leakage check mechanism so that a defective tubular battery which is determined to be leaky as a result of a check in the leakage check mechanism is rejected from the transfer disks onto a detectives collection path in order to be separated from a good-battery feed path. This arrangement allows for automatically ejecting, out of the feed path, those defective tubular batteries that have been determined to be leaky in the process of continual checks for leakage, thereby eliminating the need for providing a screening step that is carried out in accordance with the results of inspections of tubular batteries after the leakage check process. Furthermore, an alkaline battery made up of an electrolyte containing a potassium hydroxide solution may be checked to determine occurrence of leakage based on whether a fluorescent X-ray associated with a potassium component is contained in the fluorescent X-ray incident upon the fluorescent X-ray detector. Here, the potassium that is apt to emit a fluorescent X-ray is defined as the object to be checked. Thus, when applied to the detection of leakage from a battery employing an alkaline electrolyte, this arrangement allows for relying on the strength of a component associated with the potassium contained in the fluorescent X-ray to detect occurrence of leakage with high detection accuracy. Furthermore, the mask formed of a metal that does not transmit an X-ray may be allowed to condense an X-ray emitted from the X-ray source into a beam, and the beam may be then transmitted through the detection window on the housing to the sealed end face of the tubular batteries being fed. At least a length of the detection window in the direction of feed of the tubular batteries may be made variable, thereby allowing the size of the opening of the detection window to be variably adjusted corresponding to the diameter or the outer shape of a tubular battery to be checked. This makes it possible to readily perform a leakage inspection on various types of tubular batteries such as cylindrical batteries having different diameters or prismatic batteries having different outer shapes. Now, a method for checking for leakage from tubular batteries according to the present invention will be described in more detail in accordance with the embodiments with reference to the drawings. To begin with, a tubular battery will be described which is checked by a leakage check method of the present invention. FIG. 7A is a partially cut away front view illustrating an example of a tubular battery Ba to be checked that employs an alkaline electrolyte. The tubular battery Ba is configured such that a mixed positive electrode 2 and a gel zinc negative electrode 3, separated from each other by the intervention of a separator 4, are housed in a bottomed cylindrical metallic battery case 1 in conjunction with an electrolyte (not shown). Additionally, the tip of an electron collector 7 inserted in the gel zinc negative electrode 3 is disposed at the opening portion of the battery case 1, and the opening portion of the battery case 1 is sealed with an insulating gasket 8, a washer 9, and a negative electrode terminal plate 10. The opening portion of the battery case 1 of the aforementioned tubular battery Ba is sealed as follows. That is, the opening rim portion of the battery case 1 is crimped inwardly with the mutually overlapped peripheral edge portion of each of the washer 9 and the negative electrode terminal plate 10 being sandwiched by the insulating gasket 8. This causes the insulating gasket 8 to be compressively deformed, thereby providing hermeticity between the battery case 1, the insulating gasket 8, the negative electrode terminal plate 10, and the washer 9. With the tubular battery Ba, there is a possibility that the electrolyte may slightly leak through the interface between the opening rim portion of the battery case 1 and the insulating gasket 8 or between the negative electrode terminal plate 10 and the insulating gasket 8. Additionally, as shown in FIG. 7B or an enlarged view of portion VIIB of FIG. 7A, an insulating resin 11 is applied as illustrated between the peripheral opening end of the battery case 1, the end face of the insulating gasket 8, and the negative electrode terminal plate 10. Alternatively, instead of the insulating resin 11, an insulating ring may also be fitted therein. However, in any of these cases, it is intended to prevent electric short circuits between the opening rim portion of the battery case 1 and the negative electrode terminal plate 10. Note that the insulating gasket 8 may also be disposed to allow its end face to protrude from the opening rim portion of the battery case 1. However, for example, in the manufacturing process of the tubular battery Ba, a trace amount of electrolyte may be deposited on any one of the negative electrode terminal plate 10t the insulating gasket 8, or the battery case 1. Even the trace amount of electrode causes the interface between two of the members, on which the electrolyte is deposited, to be wetted with the electrolyte. This may result in a leakage path being developed therebetween, causing the electrolyte to leak along the leakage path. Such a leakage cannot be visually identified from outside due to the intervention of the aforementioned insulating resin 11 or the like. The leakage check method of the present invention is intended to ensure, by X-ray fluorescence analysis, the detection of occurrence of such leakage that cannot be visually identified. FIGS. 1 and 2 are a schematic front view and a schematic perspective view illustrating a leakage check system which embodies a method for checking for leakage from tubular batteries according to the present invention. As shown in FIG. 2, the leakage check system has a leakage check mechanism 12 for determining by X-ray fluorescence analysis whether leakage has occurred from the tubular battery Ba. The leakage check mechanism 12 includes a fluorescent X-ray detection portion 13 and an analysis portion 14. The tubular batteries Ba or an object to be checked are fed on a supply conveyor 17. When transferred from the supply conveyor 17 to a supply transfer disk 18, the tubular batteries Ba are held on the supply transfer disk 18 with their respective axial centers arranged to be parallel to each other at regular intervals. Furthermore, when the tubular batteries Ba are transferred from the supply transfer disk 18 to a main transfer disk 20 to pass through an alignment mechanism 19 of FIG. 1, the position in the orientation of their respective axial centers is realigned so that their sealed end faces are flush with each other for passage through the leakage check mechanism 12. Then, those tubular batteries Ba that have successfully passed the check at the leakage check mechanism 12 are transferred to an unloading transfer disk 21 and then fed on an unloading conveyor 22 to the next process. On the other hand, those tubular batteries Ba that have failed the aforementioned check are pushed out of the main transfer disk 20 by an ejection cylinder 23 being driven when the tubular batteries Ba are fed to a defectives ejection position as the main transfer disk 20 rotates, and thereby transferred to an ejection transfer disk 24. Thereafter, the tubular batteries Ba are conveyed on an ejection conveyor 27 to be ejected to a defectives bin or the like. Furthermore, each of the transfer disks 18, 20, 21, and 24 is designed to hold the tubular batteries Ba, which are fitted in the retaining grooves, by magnetic holding means, chucking means, or vacuuming means so that the tubular batteries Ba are not easily displaced. Accordingly, unlike a case where the tubular batteries Ba disposed upright are fed on a conveyor, there is no possibility of the tubular batteries Ba toppling over or dropping off, and thus the speed of feed can be significantly increased. Since each of the transfer disks 18, 20, 21, and 24 is disposed to rotate in a vertical plane, it is also possible to reduce the footprint of the system, thereby providing a high degree of flexibility in installation. FIG. 3 is a schematic right-side view illustrating the aforementioned leakage check system. In the figure, an apparatus casing 28 of the leakage check system is provided, on its front 28a, with each of the transfer disks 18, 20, 21, and 24 mentioned above, and the drive mechanism and the drive control mechanism for each of these transfer disks 18, 20, 21, and 24 are installed inside the apparatus casing 28. A support mount 29 is securely fixed to a lower portion on the front 28a of the apparatus casing 28. A housing 30 of the aforementioned fluorescent X-ray detection portion 13 is installed on top of the support mount 29 so that a detection window 35 provided on the housing 30 is aligned opposite to the main transfer disk 20 on the front 28a of the apparatus casing 28. Furthermore, as shown in FIG. 4, the alignment mechanism 19 shown in FIG. 1 includes a follower guide body 31 which causes each of the tubular batteries Ba, which are held at regular intervals on the main transfer disk 20 and fed in a direction of feed P, to be guided and sequentially aligned with the fluorescent X-ray detection portion 13. The alignment mechanism 19 further includes an alignment guide body 32 for abutting and thereby aligning the sealed end faces (the upper end faces in the figure) of the tubular batteries Ba, which are being fed with their orientation slightly changed by the follower guide body 31, so that the sealed end faces are flush with each other. With this arrangement, each of the tubular batteries Ba, which are held on the main transfer disk 20 with their respective sealed end faces aligned to be flush with each other, is fed at a predetermined fixed distance from the detection window 35 on the housing 30 of the fluorescent X-ray detection portion 13. FIG. 5 is a schematic view illustrating the configuration of the aforementioned leakage check mechanism 12. As also shown in FIG. 2, the leakage check mechanism 12 includes the fluorescent X-ray detection portion 13 and the analysis portion 14. In the housing 30, the fluorescent X-ray detection portion 13 includes an X-ray tube 37 serving as an X-ray source for irradiating the sealed end face 33 of the tubular batteries Ba with a primary X-ray 34 while the tubular batteries Ba are being securely carried on the aforementioned main transfer disk 20. Also included are a mask 38 for transmitting the primary X-ray 34 as a condensed beam through the detection window 35 on the housing 30, and a fluorescent X-ray detector 39 for receiving, through the detection window 35, the fluorescent X-ray (secondary X-ray) 40 that is emitted from the sealed end face 33 of the tubular batteries Ba when irradiated with the primary X-ray 34. The aforementioned mask 38 is formed of a metal such as brass that does not transmit the X-ray 34. Furthermore, the inside of the housing 30 is kept in an atmosphere of a helium gas 41, and the detection window 35 is accordingly sealed with a sealing member (not shown) made of a material such as PET film that transmits the X-rays 34 and 40. Furthermore, the housing 30 is provided with an opening control member 42, which does not transmit X-ray, for varying the shape of the opening of the detection window 35 as desired. On the other hand, the analysis portion 14 includes a detection portion 43, a computation portion 44, and a determination portion 47. The detection portion 43 detects only such a fluorescent X-ray 40 that has a wavelength associated with a predetermined component (element) among the fluorescent X-rays 40 incident upon the fluorescent X-ray detector 39. The computation portion 44 divides the fluorescent X-ray 40 detected at the detection portion 43 by the time required for a single tubular battery Ba to completely pass by its opposite area on the detection window 35, thereby determining the strength of the fluorescent X-ray 40 per unit time. The determination portion 47 compares the strength determined by the computation portion 44 with a pre-set level to determine whether leakage occurs from the tubular battery Ba. An element or one of the components of the electrolyte is pre-defined which is not used in any other parts of the tubular battery Ba and which emits an intense fluorescent X-ray. The aforementioned detection portion 43 detects the fluorescent X-ray 40 that the element emits at its wavelength. For example, suppose that an object to be checked is a tubular battery Ba that employs an alkaline electrolyte. In this case, the detection portion 43 detects the potassium in the electrolyte of a potassium hydroxide solution. Instead of computing the aforementioned strength of the fluorescent X-ray 40 per unit time, the aforementioned computation portion 44 may also divide the fluorescent X-ray 40 detected at the detection portion 43 by the surface area of the sealed end face 33 of a single tubular battery Ba to determine the strength of the fluorescent X-ray 40 per unit area. The aforementioned determination portion 47 is adapted to determine that leakage has occurred when the strength of the fluorescent X-ray 40 per unit time or the strength of the fluorescent X-ray 40 per unit area computed by the computation portion 44 is above a quantitative analysis value. The quantitative analysis value, which is experimentally pre-determined and stored as a setting, is indicative of the minimum quantity above which occurrence of leakage can be determined by a conventional manual visual inspection. FIGS. 6A to 6C are explanatory views illustrating the relative relationship between the arrangement of detection windows 35A to 35C having different shapes and respective tubular batteries Ba1 and Ba2 and the shape of a sealed end face 33 in the fluorescent X-ray check apparatus 13. FIG. 6A shows the detection window 35A having a circular opening for a cylindrical battery Ba1. FIG. 6B shows the detection window 35B having a rectangular opening for the cylindrical battery Ba1. FIG. 6C shows the detection window 35C having a rectangular opening for the prismatic battery Ba2. The detection window 35A shown in FIG. 6A for the cylindrical batteries Ba1 is formed in the shape of a circular opening such that its length in a direction of feed (here, its diameter since it is circular) L1 is less than a spacing C1 between the cylindrical batteries Ba1 being fed. In addition to this, its length in an orientation orthogonal to the direction of feed (here, also its diameter) L2 is greater than an outer size R1 of the cross-sectional shape of the cylindrical batteries Ba1 in an orientation orthogonal to their axial center (here, their diameter since they are circular). The detection window 358 shown in FIG. 6B for the cylindrical batteries Ba1 is formed in the shape of a rectangular opening such that its length L3 in the direction of feed is less than the spacing C1 between the cylindrical batteries Ba1 being fed. In addition to this, its length L4 in an orientation orthogonal to the direction of feed is greater than the outer size R1 of the cross-sectional shape of the cylindrical batteries Ba1 in an orientation orthogonal to their axial center. The detection window 35C shown in FIG. 6C for the prismatic batteries Ba2 is formed in the shape of a rectangular opening such that its length L5 in the direction of feed is less than a spacing C2 between the prismatic batteries Ba2 being fed. In addition to this, its length L6 in an orientation orthogonal to the direction of feed is greater than an outer size R2 of the cross-sectional shape of the prismatic batteries Ba2 in an orientation orthogonal to their axial center. A description will now be made to the inspection process in the aforementioned leakage check system that embodies a leakage check method of the present invention. As shown in FIGS. 1 and 2, the tubular batteries Ba having been fabricated to serve as a battery are transferred to the main transfer disk 20 via the supply conveyor 17 and the supply transfer disk 18, to be thereby aligned and held in parallel to each other at predetermined intervals on the main transfer disk 20. Then, in a step where they are fed to the leakage check mechanism 12 as the main transfer disk 20 rotates, the alignment mechanism 19 aligns their positions in the orientation of their axial center so that their respective sealed end faces 33 are flush with each other, and thereafter they are kept as aligned. Therefore, as shown in FIG. 5, each of the tubular batteries Ba is allowed to pass through the leakage check mechanism 12 while each sealed end face 33 is always kept at a fixed constant distance D to the detection window 35. Accordingly, the distance D from the sealed end face 33 to the detection window 35 can be set to a very small value, for example, approximately 2 mm. When each of the tubular batteries Ba passes through the leakage check section that opposes the detection window 35, the primary X-ray 34 emitted from the X-ray tube 37 is condensed through the mask 38 into a beam, which passes through the detection window 35 to irradiate the sealed end face 33 of a tubular battery. Then, the fluorescent X-ray 40 coming out of the sealed end face 33 enters the fluorescent X-ray detector 39 through the detection window 35. In the analysis portion 14, among the fluorescent X-rays 40 that are incident upon and detected by the fluorescent X-ray detector 39, the detection portion 43 detects only such a fluorescent X-ray 40 that has a wavelength associated with a predefined one of those components contained in the electrolyte. As the predefined component, it is preferable to define an element of the components of the electrolyte which is not used in any other parts of the tubular batteries Ba and emits an intense fluorescent X-ray 40. For example, when an object to be checked is a tubular battery Ba that employs an alkaline electrolyte, it is preferable to define potassium in the electrolyte of a potassium hydroxide solution. The computation portion 44 of the analysis portion 14 has the following various types of data pre-stored therein. That is, the computation portion 44 has at least the following time settings stored therein. These settings include: the timing at which the end portion of each of the tubular batteries Ba in the direction of feed is brought to oppose the detection window 35; the time required for a tubular battery Ba to completely pass by the detection window 35 when the main transfer disk 20 is continuously driven at a constant rotational speed and driven at a varied low speed when the tubular battery Ba is passing by the detection window 35 opposing thereto; the standstill time during which the main transfer disk 20 is intermittently driven to allow a tubular battery Ba to temporarily stand still opposing the detection window 35; and the surface area of the sealed end face 33 of various types of tubular batteries Ba to be checked. Furthermore, the aforementioned computation portion 44 divides the fluorescent X-ray 40, which is detected at the detection portion 43 from the predefined timing point until any one of the aforementioned time, settings has elapsed, by the time required for a single tubular battery Ba to completely pass by the detection window 35. In this manner, the strength of the fluorescent X-ray 40 per unit time is determined. Alternatively, the aforementioned computation portion 44 divides the fluorescent X-ray 40, which is detected at the detection portion 43 from the predefined timing point until any one of the aforementioned time settings has elapsed, by the surface area of the sealed end face 33 of a tubular battery Ba being inspected. In this manner, the strength of the fluorescent X-ray 40 per unit area is determined. Subsequently, the determination portion 47 compares the strength determined by the computation portion 44 with the predefined level which has been experimentally determined through a manual visual inspection as described above. Then, when the determined strength is above the level setting, the determination portion 47 determines that leakage has occurred in the tubular battery Ba. As explained in relation to FIGS. 6A to 6C, in determining occurrence of leakage by X-ray fluorescence analysis as described above, any one of the circular detection window 35A or the rectangular detection windows 35B and 35C may be employed as the detection window 35. In any case, the length L1, L3, and L5 of the detection windows 35A to 35C in the direction of feed of the tubular batteries Ba1 and Ba2 are defined to be less than the spacing C1 or C2 between the tubular batteries Ba1 or Ba2. Accordingly, even when each of the tubular batteries Ba1 and Ba2 arranged at the smallest possible spacing C1 or C2 is fed at high speeds, the fluorescent X-rays 40 from two adjacent ones of the respective tubular batteries Ba1 and Ba2 will never simultaneously enter the fluorescent X-ray detector 39 through the detection window 35B or 35C. Accordingly, the fluorescent X-rays 40 emitted individually from the tubular batteries Ba1 and Ba2 and received by the fluorescent X-ray detector 39 can be separately identified, thereby allowing for dramatically increasing the speed of checking for leakage. As a result of actual measurements, it was ensured that 800 to 1200 tubular batteries Ba were checked per one minute, and the inspection speed can be further increased to such an extent as to check 2000 tubular batteries Ba per one minute. Furthermore, the lengths L2, L4, and L6 of the detection windows 35A to 35C in an orientation orthogonal to the direction of feed are defined to be slightly greater than the outer sizes R1 and R2 of the cross-sectional shape of the tubular batteries Ba1 and Ba2 in an orientation orthogonal to their axial center. This ensures that the fluorescent X-ray 40 emitted from any part of the sealed end face 33 of the tubular batteries Ba1 and Ba2 enters the fluorescent X-ray detector 39, thereby allowing for detecting occurrence of leakage even when the leakage occurs at any part of the sealed end face 33. Furthermore, the aforementioned lengths L2, L4, and L6 are defined to be slightly larger than the outer sizes R1 and R2 of the aforementioned cross-sectional shapes, thereby allowing the detection windows 35A to 35C to block the entry of a fluorescent X-ray from the surrounding of the tubular batteries Ba1 and Ba2 into the fluorescent X-ray detector 39. It is thus possible for the detection portion 43 of the analysis portion 14 to detect, at a good S/N ratio, the fluorescent X-ray 40 having a predetermined wavelength among the fluorescent X-rays 40 that are incident upon the fluorescent X-ray detector 39. Furthermore, the computation portion 44 and the determination portion 47 of the analysis portion 14 determine occurrence of leakage by comparing either one of the following intensities with a level setting that is experimentally pre-determined. That is, the intensities include the strength of the fluorescent X-ray 40 per unit time that is determined by dividing the fluorescent X-ray 40 of the predetermined wavelength, which has entered the fluorescent X-ray detector 39 when a tubular battery Ba passes by the detection window 35, by the time required for the entire tubular battery Ba to completely pass by the detection window 35. The intensities also include the strength of the fluorescent X-ray 40 per unit area that is determined by dividing the fluorescent X-ray of the predetermined wavelength, which has entered the fluorescent X-ray detector 39 when the entire tubular battery Ba passes by the detection window 35, by a cross-sectional area corresponding to the surface area of the sealed end face 33 of the tubular battery Ba. This makes it possible to highly accurately detect occurrence of leakage no matter how the tubular batteries Ba are fed, for example, continuously at a constant speed, intermittently at a standstill when a tubular battery Ba opposes the detection window 35, or variably at a low speed when a tubular battery Ba passes by the detection window 35. Furthermore, the inside of the housing 30 of the fluorescent X-ray detection portion 13 is kept in an atmosphere of the helium gas 41, and the helium gas 41 reduces the argon gas contained in the air. It is thus possible to remove adverse effects exerted by the argon gas that emits a fluorescent X-ray 40 at a wavelength that is similar to that of the potassium to be detected. This allows for eliminating noise caused by the argon gas and thereby highly accurately detecting the strength of the fluorescent X-ray 40 of the predetermined wavelength at a high S/N ratio. In the aforementioned fluorescent X-ray detection portion 13, the mask 38 condenses the primary X-ray 34 into a beam, thereby allowing the opening area of the detection windows 35A to 35C to be reduced as small as possible as explained in relation to FIGS. 6A to 6C. In addition to this, the distance D (see FIG. 5) between the detection window 35A to 35C and the sealed end face 33 of the tubular batteries Ba1 and Ba2 being fed can be set to as small a value as approximately 2 mm as described above. It is thus possible to dramatically reduce the aforementioned adverse effects caused by an argon gas contained in the air that is present between the detection window 35A to 35C and the sealed end face 33 of the tubular batteries Ba1 and Ba2 being fed. This also allows for further improving the accuracy of detecting leakage. In the aforementioned leakage check system, each of tubular batteries Ba arranged in a predetermined manner is held on the main transfer disk 20 to be fed in a rotary scheme toward the fluorescent X-ray detection portion 13. Each of the tubular batteries Ba can be thus positively held so as not to be displaced out of position while being fed at a high speed, thereby ensuring that the sealed end face 33 of the tubular batteries Ba opposes and passes by the detection window 35 at the shortest possible constant distance D therebetween. This makes it possible to detect leakage with high accuracy while tubular batteries are detected at further increased speeds. Furthermore, as explained in relation to FIG. 3, the fluorescent X-ray detection portion 13 is attached to the support mount 29 so that the detection window 35 on the housing 30 is oriented rearward for the primary X-ray 34 to be emitted rearward. This arrangement eliminates the risk of the operator being exposed to the primary X-ray 34 when working in front of the support mount 29 opposite to the attachment of each of the transfer disks 18, 20, 21, and 24, thereby providing a very safe leakage check system. Furthermore, as shown in FIG. 5, the shape of the opening of the detection window 35 can be varied as desired by controlling the opening control member 42. Accordingly, when different tubular batteries Ba are to be checked, the opening control member 42 can be adjusted to accommodate the diameter or the outer shape of the tubular batteries Ba. This makes it possible to readily perform a leakage inspection on various types of tubular batteries Ba such as cylindrical batteries Ba1 having different diameters or prismatic batteries Ba2 having different outer shapes. Then, those defective tubular batteries Ba that have been determined to be leaky in the leakage inspection at the leakage check mechanism 12 are pushed out with the ejection cylinder 23 (see FIG. 1) and thereby transferred to the ejection transfer disk 24. Here, the ejection cylinder 23 is driven at a point in time at which a tubular battery Ba is fed to oppose the ejection transfer disk 24 as the main transfer disk 20 rotates. Thereafter, the tubular batteries Ba are automatically collected such as in a bin via the ejection conveyor 27 that constitutes a defectives collection path. Accordingly, the leakage check system separates defective tubular batteries Ba from good tubular batteries Ba in a continuous transport process to automatically reject those defectives. It is thus not necessary to provide a go-or-no-go screening step that is carried out in accordance with the results of leakage inspections of tubular batteries Ba after the leakage inspection step. Note that in the aforementioned embodiment, such an example has been illustrated in which the tubular batteries Ba are fed while being held in position on the transfer disks 18, 20, 21, and 24. However, the same effects as described above can also be obtained by allowing the tubular batteries Ba to be fed on a straight reed path while being held in position. As a matter of course, various additional components may also be employed when necessary for practical use. The detection window 35 may be provided with a shutter to block it to intercept the X-ray 34 when the system is stopped, thereby providing further enhanced safety. Furthermore, the main transfer disk 20 may be provided with an additional mechanism for rotating a held tubular battery Ba around itself only in a region where it passes by the detection window 35, thereby providing further improved accuracy of detecting leakage. As described above, according to the present invention, a detection window allows an X-ray to pass therethrough and then irradiate the sealed end face of tubular batteries and allows a fluorescent X-ray emitted from the sealed end face to pass therethrough to be incident upon a fluorescent X-ray detector. The length of the detection window corresponding to the direction of feed of the tubular batteries is less than the spacing between the tubular batteries being fed. The length of the detection window in an orientation orthogonal to the direction of feed is greater than the outer size of the cross-sectional shape of the tubular batteries in an orientation orthogonal to their axial center. This makes it possible to realize a method for checking for leakage from tubular batteries, which allows for accurately determining occurrence of leakage from a tubular battery at high speeds by X-ray fluorescence analysis. |
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summary | ||
058898328 | description | DETAILED DESCRIPTION PREFERRED EMBODIMENT The control cluster shown in FIG. 1 comprises a spider 10 having twenty-four rods 12 suspended therefrom. The spider comprises a hub 14 from which sixteen fins 16 are mounted to radial carrying fingers 18 in which the top end plugs of absorbent rods 12 are fixed. The fingers 18 project well away from the fins in a downward direction. A counterbore 20 is formed in the hub 14 and has grooves for receiving a device (not shown) for connection to a drive shaft for displacing the cluster vertically. In the bottom portion of the hub there is provided a bore 22 for receiving a shock absorber. The shock absorber comprises a socket 24 and a spring 25 bearing against an internal partition separating the bore from the counterbore. A screw 26 whose position is adjusted by being screwed into tapping formed through the partition serves to fix the rest position of the socket 24, in which position it projects downwards from the hub. After adjustment, the screw is locked into place by being welded to the bottom of the counterbore 20. The fins are conventionally fixed to the hub by brazing. In FIG. 2, the fingers 18 are fixed to the fins by a tenon-and-mortise joint and by brazing. The terminal portion (or intermediate portion for the inner fingers) constitutes a tenon which is engaged in a slot in the finger, which slot constitutes a mortise. In the prior art, each finger is formed with by a tapped blind hole for fixing the plug 26 of a rod 12. This blind hole cannot extend beyond the bottom edge of the fins because of the tenon-and-mortise assembly. FIG. 3 shows one possible way of fixing a rod in the tapped blind hole of a finger 18. An extension of the plug is terminated by a threaded length 29 which is screwed into a tapped portion of the hole. Between the threaded length and the plug proper 46, there is provided a portion of smaller diameter situated between two bulges. A pin or peg 27 locks the connection. FIG. 4 shows the structure of a spider constituting a first embodiment of the invention. The spider is made as a single piece that is either molded directly or obtained by electro-erosion. To make it easier to mold, the edges of the fins are constituted by flat portions that are mutually parallel, constituting steps that are interconnected by sloping flat portions. The end fingers 18a are of a height that corresponds exactly to that of the portions of the fins 16 remote from the axis of the hub. The intermediate fingers 18b have the same height as the portions of the fins that are nearer to the axis of the hub. FIG. 4 does not show the bores formed along the axes of the fingers, but it does show some of the bullet-shaped nuts 28 for fixing rod plugs. The spider shown in FIG. 4 can receive a shock absorber of the same kind as that shown in FIG. 1. Nevertheless, in order to facilitate manufacture and assembly, it is advantageous to adopt the disposition shown in FIG. 5, where members corresponding to those of FIG. 1 are given the same reference numerals. The rest position taken up by the socket 24 under drive from the spring 25 is fixed by bearing against a washer 30 mounted in an end counterbore 33 of the bore 22 and welded into place. The socket 24 has a shoulder enabling it to bear against the inside portion of the top face of the washer 30. In the embodiment shown in FIG. 6, the spider is in two parts. It comprises firstly a tubular part or sleeve 14a constituting the top portion of the hub, and secondly a one-piece bottom part comprising the bottom portion of the hub together with the fins 16 and the fingers 18. The sleeve 14a is machined. The bottom portion which includes the bottom portion of the hub, the fins, and the fingers is obtained by molding a solid part or by electro-erosion. The connection between the top portion 14a and the bottom portion can be achieved in various ways. As shown in FIG. 6, a cylindrical step is machined in the top end of the bottom portion. The bottom end of the top portion 14a has a complementary step that engages in the bottom portion. A circular weld 42 serves to hold together the two portions of the spider. The bores 22 formed in each of the two portions have the same diameter so as to avoid any discontinuity in the bore. In the modified embodiment shown in FIG. 7, the connection between the bottom end of the top portion and the bottom portion is a screw connection. For this purpose, the upper portion of the fraction of the bore situated in the bottom portion of the spider is tapped while the sleeve is threaded. Thus, the two portions of the spider can be fixed together by screw engagement until the shoulder 44 of the upper portion presses against the top end of the fins. The spider structures shown in FIGS. 6 and 7 enable the shock absorber to be assembled without any additional element for retaining the socket 24. It suffices to provide a lip 34 -at the bottom end of the bore in the bottom portion against which a shoulder formed on the socket 24 can bear. The socket 24 and the spring 25 are put into place prior to the sleeve being assembled on the bottom portion of the spider (FIG. 7). This shock absorber structure makes it possible to provide a cooling fluid flow duct through the spider. No screw is required. The cooling fluid can flow along an internal channel 36 in the socket and through a central hole 38 in the collar against which the spring 25 bears. There is no longer any need to adjust and weld a screw. The arrangement illustrated in FIG. 7A mainly differs from that of FIG. 7 in that there is a thimble 45 for locking against rotation, so as to prevent the threaded connection between the bottom part and the lower end of the top part from unthreading. The thimble 45 is in abutment against the shoulder of the top portion and has at least one finger which engages into a slot of the bottom portion. After the top portion has been screwed down, the thimble is secured by some welding points at 47. The modification shown in FIG. 7B differs from that of FIG. 7A in that the thimble 45 is replaced with a key 72 located in a recess 74 of the top portion and projecting into a slot 76 of the bottom portion. The parts are again screwed into each other and the recess 74 and slot 76 are located in front of each other. The key, which preferably has half-cylindrical ends, is slid into place as indicated by an arrow. It is then secured to the top portion by welding along its upper edge. In FIG. 7C, the smooth bore of the bottom portion has a lower counterbore for receiving a nut 78 apt to be screwed on the threaded end portion of hub 14A. After the nut has been screwed down, it may be secured by welding at points 80, through the metal, between two fins. The rods can be fixed to the spiders shown in FIGS. 4, 6, and 7-7C by means of the type described in the above-mentioned French Patent No. 2,599,884. Nevertheless, it is preferable to adopt one of the dispositions described below which give flexibility to the cluster of rods facilitating insertion into the guide tubes of an assembly or of guide sheaths in the upper internal of the reactor. In the case shown in FIG. 8, each rod 12 has a plug 46 provided with an extension having one or more portions 48 of small section, and rounded connections with adjacent portions of nominal diameter to avoid breakage starters. There may be up to three generally-cylindrical portions of small section. The end portion 50 of the extension may also be fixed to the respective finger by means of the kind described in the above-mentioned French patent document. In the variant shown in FIG. 9, the fixing means still comprise an extension of the plug 46. The extension has a portion 54 of small diameter located between two swellings. It is fixed in a connector 52. The connector 52 has two cylindrical bearing surfaces 56 on either side of the length of small diameter. Only the bearing surfaces bear against the wall of an open bore formed in the finger 18. A flat 57 in the bore receives a flat on the connector 52 so as to prevent the connector from rotating. The top bearing surface 56 extends beyond the upper surface of the finger and guides a cap shaped nut 58 screwed onto the threaded terminal portion of the extender. When the cap nut is tightened, it presses a shoulder of the extender 52 against the bottom edge of the finger. Once the nut has been fully tightened, it can be prevented from rotating by deforming a thin ferrule of the kind described in French Patent No. 2,599,884. In the variant shown in FIGS. 10 and 11, the plug 46 of the rod is fixed to the finger 18 by an extension of a structure such that the rod is slightly loose. The extension has a threaded terminal portion for receiving a cap similar to that of FIG. 9. A pin or peg 61 prevents the plug from rotating. A groove 68 may be formed in the extension to enable it to be engaged when the pin is in place. The extension also presents a portion of small section situated in the hole of the finger and lying between two projections, swellings or bulges 62 and 64. The top swelling 62 has a sliding fit in the hole and has a centering function. The diameter of the bottom swelling 64 is slightly smaller than that of the hole so as to allow the rod a limited amount of lateral movement. An internal shoulder 66 may be provided in the hole for receiving the extension to constitute a bearing surface for the top face of the swelling 62. The embodiment shown in FIG. 12 differs from that of FIG. 10 essentially in that the projection 62 is connected to a portion of smaller diameter via a frustoconical zone 70 that co-operates with a corresponding portion of the hole to hold the rod in translation. The rod is prevented from rotating during assembly or disassembly by tooling (not shown). A portion of reduced section and/or a set of bottom swellings also ensure that the rod can move laterally. The modified embodiment illustrated in FIG. 12A is designed for providing an accurate degree of centering of the extension of the fuel rod into the finger 18 and at the same time to prevent jamming due to tilting when the fuel rod is mounted or dismounted. For that, the projection 62 has a very short cylindrical portion (lesser than 1 mm) extended upwardly by a frusto-conical abutment zone 70, having an apex angle of about 110.degree., and extending downwardly as a frusto-conical zone 82 merging with the portion of reduced diameter. FIG. 13 and 14 show a possible distribution for locking pegs or pins 61 used in the embodiment of FIG. 10. This distribution gives easy access to all of the pegs 61 for installation or removal purposes. |
summary | ||
claims | 1. A radioprotective product comprising a radioprotective material comprising a polyamino acid or polyamino acids or at least one derivative selected from the group consisting of polyamino acid esters, and copolymers comprising polyamino acids and other polymers (except for polyamino acid compounds in which a residue of aminobenzaldehyde semicarbazone or aminobenzaldehyde thiosemicarbazone is bonded to carboxylic acid at the ω-position of polyglutamic acid or polyaspartic acid). 2. A radioprotective product according to claim 1, which is at least one member selected from the group consisting of radioprotective films, radioprotective sheets, radioprotective coating agents, radioprotective cosmetic products, radioprotective fiber products, radioprotective preparations and radiation-resistant medical materials. 3. A method for producing a radioprotective product, comprising:administering or admixing a polyamino acid or polyamino acids or at least one derivative selected from the group consisting of polyamino acid esters, and copolymers comprising polyamino acids and other polymers (except for polyamino acid compounds in which a residue of aminobenzaldehyde semicarbazone or aminobenzaldehyde thiosemicarbazone is bonded to carboxylic acid at the ω-position of polyglutamic acid or polyaspartic acid), to or with a product in an amount effective to make said product radioprotective. 4. A method for blocking or reducing adverse effects of radiations on an object, the method, comprising contacting the object with a radioprotective material comprising a polyamino acid or polyamino acids or at least one derivative selected from the group consisting of polyamino acid esters, and copolymers comprising polyamino acids and other polymers (except for polyamino acid compounds in which a residue of aminobenzaldehyde semicarbazone or aminobenzaldehyde thiosemicarbazone is bonded to carboxylic acid at the ω-position of polyglutamic acid or polyaspartic acid). 5. A method according to claim 4, wherein the radioprotective material is mixed with, placed over, or immobilized on the object. 6. A method for blocking or reducing adverse effects of radiations on an object, the method comprising contacting the object with the radioprotective product according to claim 1. 7. A method according to claim 6, wherein the radioprotective product is administered to, placed over or applied to the object. 8. A method for production of a radioprotective material, comprising:administering or admixing a polyamino acid or polyaminoacids or at least one derivative selected from the group consisting of polyamino acid esters, and copolymers comprising polyamino acids and other polymers, for production of a radioprotective product which provides protection from at least one member selected from the group consisting of cosmic rays, radio waves, electromagnetic waves, infrared radiation, visible light, alpha rays, beta rays, proton beams, baryon beams, X rays, gamma rays, electron beams and neutron beams. 9. A method of producing a radioprotective product comprising using a radioprotective material produced by a method comprising adding a polyamino acid or acids or a derivative or derivatives thereof to a base material, wherein if the radioprotective material is a polyamino acid compound having a residue of aminobenzaldehyde semicarbazone or aminobenzaldehyde thiosemicarbazone, the residue is not bonded to carboxylic acid at the ω-position of polyglutamic acid or polyaspartic acid. |
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claims | 1. A detection system configured to provide for increased penetration of an object, comprising:a source configured to generate radiation directed to an inspection volume;a collimator positioned between the source and the object, wherein the collimator is configured to receive the radiation and comprises two or more actuators and two or more attenuators, and wherein each of the two or more actuators is coupled with one of the two or more attenuators to cause each of the two or more attenuators to not attenuate, attenuate, or partially attenuate the radiation, thereby producing one or more fanlets from the radiation;a detector array opposing said source and positioned within the inspection volume for detecting the one or more fanlets projected on and through the object;a controller configured to synchronize the source and the collimator and collect data from the detector array corresponding to each of the one more fanlets;a conveyor adapted to move the object through the inspection volume, wherein the controller is configured to control the conveyor such that a total time for the object to pass through each of the one or more fanlets multiplied by a rate of speed of the conveyor is equal to or less than a width of a detector in the detector array; anda processing unit for combining the collected data into a composite image. 2. The detection system of claim 1, wherein each of the one or more fanlets has an angular range greater than 1 degree but smaller than the angular coverage of the object. 3. The detection system of claim 1, wherein the data from the detector array corresponding to each of the one more fanlets is representative of an image slice of the object and wherein the processing unit combines the image slices into the composite image. 4. The detection system of claim 1, wherein the source is a pulsed X-ray source configured to generate interlaced dual energy beams. 5. The detection system of claim 1, wherein the X-ray source is a pulsed x-ray source configured to produce-X-ray pulses comprising low and high energy X-ray beams separated in time. 6. The detection system of claim 1, wherein the collimator is configured to generate an overlap between each of the one or more fanlets. 7. The detection system of claim 1, wherein the one or more fanlets move vertically. 8. The detection system of claim 1, wherein the processing unit is configured to collect image slices from the detector array corresponding to a complete scan cycle of the one or more fanlets. 9. The detection system of claim 1, wherein the controller is configured to adjust at least one of a beam intensity or energy of each of the one or more fanlets based on signals detected from a previous fanlet at a same vertical position with respect to the object to cause each vertical position to be subject to interlaced dual energy scanning. 10. The detection system of claim 1, wherein each of the two or more actuators is physically coupled by a member to one of the two or more attenuators. 11. The detection system of claim 1, wherein the two or more actuators are rotary actuators. 12. The detection system of claim 1, wherein the collimator is configured in a vertical position and wherein the collimator projects a fan beam that covers the vertical extent of the object being scanned. 13. A detection system configured to provide for increased penetration of an object, comprising:a source configured to generate radiation directed to an inspection volume;a collimator positioned between the source and the object, wherein the collimator is configured to receive the radiation and comprises two or more actuators and two or more attenuators, and wherein each of the two or more actuators is coupled with one of the two or more attenuators to cause each of the two or more attenuators to not attenuate, attenuate, or partially attenuate the radiation, thereby producing one or more fanlets from the radiation;a detector array opposing said source and positioned within the inspection volume for detecting the one or more fanlets projected on and through the object;a controller configured to synchronize the source and the collimator and collect data from the detector array corresponding to each of the one more fanlets, wherein the controller is configured to adjust at least one of a beam intensity or energy of each of the one or more fanlets based on signals detected from a previous fanlet at a same vertical position with respect to the object; anda processing unit for combining the collected data into a composite image. 14. The detection system of claim 13, wherein each of the one or more fanlets has an angular range greater than 1 degree but smaller than the angular coverage of the object. 15. The detection system of claim 13, wherein the data from the detector array corresponding to each of the one more fanlets is representative of an image slice of the object and wherein the processing unit combines the image slices into the composite image. 16. The detection system of claim 13, wherein the source is a pulsed X-ray source configured to generate interlaced dual energy beams. 17. The detection system of claim 13, wherein the X-ray source is a pulsed x-ray source configured to produce-X-ray pulses comprising low and high energy X-ray beams separated in time. 18. The detection system of claim 13, wherein the collimator is configured to generate an overlap between each of the one or more fanlets. 19. The detection system of claim 13, wherein the processing unit is configured to collect image slices from the detector array corresponding to a complete scan cycle of the one or more fanlets. 20. The detection system of claim 13, wherein adjusting at least one of a beam intensity or energy of each of the one or more fanlets based on signals detected from a previous fanlet at a same vertical position with respect to the object causes each vertical position to be subject to interlaced dual energy scanning. 21. The detection system of claim 13, wherein each of the two or more actuators is physically coupled by a member to one of the two or more attenuators. 22. The detection system of claim 13, wherein the two or more actuators are rotary actuators. 23. The detection system of claim 13, wherein the collimator is configured in a vertical position and wherein the collimator projects a fan beam that covers the vertical extent of the object being scanned. |
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047327294 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention was made resulting from the study of the structure of a conventional fast breeder reactor. FIG. 1 shows the reactor core of a conventional fast breeder reactor. Below the core is formed a high-pressure plenum 7 defined by a cylinder 10, an upper supporting plate 3 and a lower supporting plate 4. Below the lower supporting plate 4, there is a medium-pressure plenum 8 formed by a plate 11 carried by the cylinder 10. Each sleeve 5 having openings 6 is installed between the upper supporting plate 3 and the lower supporting plate 4. The entrance nozzle 15 of a fuel assembly 2 and the lower end portion of a control rod guide pipe are inserted into the sleeves 5 (Refer to FIG. 2). The cylinder 10 is supported in a reactor vessel by means of a core-supporting structure (not shown). The core-supporting structure separates the inside space of the reactor vessel into an upper plenum 12 and a lower plenum 13. Sodium in the lower plenum 13 is sucked by a main circulating pump, which is not shown and which boosts up the pressure, of the sodium and is delived into the high-pressure plenum 7 through a delivery nozzle 14. The sodium from the high-pressure plenum 7 is introduced into each fuel assembly 2 and each control rod guide pipe via openings 6. The sodium is introduced into each fuel assembly 2 through orifices 16 in an entrance nozzle 15. It is noted that the pressure of the medium-pressure plenum 8 is set lower than that of the high-pressure plenum 7 but slightly higher than that of the upper plenum 12 (low-pressure plenum). The medium-pressure plenum 8 is provided to check the floating of the fuel assemblies 2. In other words, in FIG. 2, since the pressure loss of the bundle portion of the fuel assembly 2, .DELTA.P=P.sub.A -P.sub.B, is in a range of 3-5 kg/cm.sub.2, the fuel assembly is subjected to an upward thrust force. Assuming that the area within the wrapper tube of the fuel assembly 2 is Aai, the magnitude of this thrust force is given by the formula, Aai.multidot.(P.sub.A -P.sub.B). To counteract this thrust force, a downward thrust force is applied preferably to the fuel assemblies. This can be obtained by the medium-pressure plenum 8 which gives a downward thrust force of A.sub.N .multidot.(P.sub.A -P.sub.L) where A.sub.N is the pressure-receiving area of the entrance nozzle 15. Thus it becomes possible to reduce the floating force of the fuel assemblies 2. In a case where a coolant influx system as is shown in FIG. 2 is adopted, the coolant in the high-pressure plenum 7 flows in such a manner that it crosses the group of entrance nozzles of several hundred fuel assemblies 2 densely arranged. Hence, a large radial pressure loss results within the high-pressure plenum 7. For this reason, it is liable to be impossible to maintain a sufficient influx pressure. In order to eliminate this problem, conventionally the length of the entrance nozzles 15 is usually made longer, and as well the height (H.sub.P) of the high-pressure plenum is made higher to make the flow rate of the coolant low, thereby decreasing the pressure loss of the flow. Next, it is conceivable that a bending moment may act on an upper spherical seat for each fuel assembly 2, which is located on the upper fitting portion of the entrance nozzle 15, owing to the horizontal vibration of the fuel assembly 2 as a result of an earthquake or the like. Therefore, the entrance nozzle 15 is required to withstand this bending stress. However, at the portion of the entrance nozzle 15 where the orifice 16 is formed, the rigidity of the pipe is comparatively low. For this reason, the length M (refer to FIG. 2) of the upper fitting portion of the entrance nozzle 15 is extended, thereby lowering the entrance nozzle 15 to make smaller the bending moment applied to the entrance nozzle 15. This problem could be solved by thickening the wall thickness of the entrance nozzle 15. Yet, making smaller the inside diameter of the entrance nozzle would cause the flow rate of the coolant to increase. On the other hand, making larger the outside diameter of the entrance nozzle 15 would cause the intervals of the sleeves 5 to become too small. Hence, it would be impossible to obtain the desired pressure-reducing characteristics of the orifice. Because of the pressure loss of the coolant and the restriction in aseismatic strength, as explained above, the length of the entrance nozzle 15 becomes extremely long. In addition, in a fast breeder reactor like the one explained above, the amount of heat generated in the control rod is small, so that, in terms of the flow rate of the coolant (sodium), only about 10% of that for the fuel assembly may be used. However, since there is a very small amount of heat generation due to (n, .alpha.) reaction, it is not feasible to completely dispense with cooling. The channel resistance in the control rod is relatively small as compared with that for the fuel assembly. In particular, a cylinderical annular gap between the control rod guide pipe and the control rod is made substantially large (approximately 1 cm) in order to enhance the reliability of the insertion of the control rod at the time of an earthquake. Consequently, a large quantity of sodium can flow under a very small pressure differential. Accordingly, in the case of supplying the sodium in the high-pressure plenum 7 into the control rod guide pipe, each sleeve 5 into which the control rod guide pipe is inserted, and the sodium influx holes in the lower portion of the control rod guide pipe must be made relatively small, and, at the same time, a multi-stage orifice structure must be adopted in which the diameter of the posterior stage of the influx hole is made smaller than that of the preceding stage thereof. In such a case, however, the internal pressure loss between the high-pressure plenum 7 and the control rod guide pipe increases, which impedes the insertion of the control rods at the time when the control rods are rapidly inserted into the reactor core during a scram. At the time of rapidly inserting the control rods, it is necessary to discharge the sodium located below the control rod within the control rod guide pipe outside of the guide pipe. If the aforementioned orifice is small, however, discharge to the high-pressure plenum is restricted. Consequently, this results in an increase in the quantity of the sodium which passes through the gap between the control rod and the control rod guide pipe and through the inside of the control, and is discharged upward, causing the lowering of the control rod to be retarded. As a result of conducting examinations to solve these problems, the inventors found out that these problems can be solved by disposing the medium-pressure plenum in the top portion and the high-pressure plenum in the lower portion and by supplying into the fuel assemblies the coolant from the high-pressure plenum through the lower side, and as well supplying into the control rod guide pipe the coolant from the medium-pressure plenum in the radial direction. An embodiment of the present invention will be described hereafter. A preferred embodiment of the present invention applied to a tank-type fast breeder reactor is shown in FIGS. 3 and 4. The tank-type fast breeder reactor has a reactor vessel 21 installed to a roof slab (not shown) supported by the building. An intermediate heat exchanger 22, a main circulating pump 23, and the upper structure 24 of the core are installed on the roof slab, within the reactor vessel 21. A guard vessel 25 surrounds the reactor vessel 21. A cylinder 10 is installed in the reactor vessel 21 by means of a core-supporting structure 26. An upper supporting plate 3, a lower supporting plate 4 and a plate 11 are installed in the cylinder 10 downward in the mentioned order. FIG. 4 shows a transverse sectional view of a reactor core 1. The reactor core 1 comprises a core region where core fuel assemblies 2a are disposed in its center, and a blanket region where blanket fuel assemblies 2B are disposed, in the peripheral portion surrounding the core fuel assemblies 2A. Each core fuel assembly 2A has a multiplicity of fuel pins which contain plutonium, i.e., a fissionable material. The fuel pins in the blanket fuel assemblies 2B mainly contain natural uranium or depleted uranium. The control rod 27 move vertically in the core region among the fuel assemblies. The control rods 27 are coupled with a control rod drive unit 28 installed on the upper structure of the core and are moved vertically by the same. The core fuel assemblies 2A and the blanket fuel assemblies 2B are supported by the lower supporting plate 4. Detailed description will be made to the supporting of each core fuel assembly 2A with reference to FIG. 5. The supporting of each blanket fuel assembly 2B is also effected in a similar manner. FIG. 5 shows the lower end portion of the core fuel assembly 2A. An opening 30 into which the entrance nozzle (diameter: D.sub.2) of the core fuel assembly 2A is inserted is provided in the upper supporting plate 3. In the lower supporting plate 4, a stepped opening 31 for preventing misloading of the core fuel assembly 2A is provided, and a through hole 32 with a diameter d.sub.3 through which the sodium flows is also provided. An annular bead 33 with a radius R is provided at the bottom of the opening 31. A foreign object removing plate 34 having small holes is disposed at the entrance side of the through hole 32, and is attached to on the lower supporting plate 4. The entrance nozzle 29, whose diameter is small at its lower end (D.sub.2 >D.sub.3), is incorporated to the lower end portion of the wrapper tube 35 of the core fuel assembly 2A. The fuel pin containing plutonium is housed in the hexagonally-shaped wrapper tube 35. A labyrinth 36 is formed on the outside of the upper end portion of the entrance nozzle 29. A channel 59 is formed in the entrance nozzle 29. This channel 59 is a hole which has a diameter d.sub.3 at its lower end portion, increasing its diameter in the upward direction from the lower end portion. A plurality of orifices 37 are disposed radially in the entrance nozzle 29. The lower end surface of the entrance nozzle 29 is recessed and is inclined inward and upward. The lower end surface of such an entrance nozzle 29 is in tangential contact with the annular bead 33. The core fuel assembly 2A is supported on a line 38 by the lower supporting plate 4. Next, description will be made of how the control rod guide pipe is supported. FIG. 6 shows the control rod guide pipe 39. The control rod guide pipe 39 comprises a hexagonally-shaped cylindrical body 40, and the entrance nozzle 41 which is installed at the lower end portion thereof. Openings 42 are formed in the entrance nozzle 41. A dashpot 43 is placed in the control rod guide pipe 39. The upper and lower ends of the sleeve 45 having openings 44 are installed on the upper supporting plate 3 and the lower supporting plate 4, respectively. The control rod guide pipes 40 are arranged corresponding to the positions of the control rods 27, as shown in FIG. 4. Each control rod 27 coupled with the control rod drive unit 28 is disposed in the control rod guide pipe 39. A dash ram 46 which is located at the lower end of the control rod 27 is inserted into the dashpot 43 when the control rod 27 has been completely inserted into the reactor core. The dashpot 43 and the dash ram 46 serve as a damper at the time of rapid insertion of the control rod 27. In FIG. 3, a medium-pressure plenum 47 is defined between the upper supporting plate 3 and the lower supporting plate 4, while a high-pressure plenum 48 is defined between the lower supporting plate 4 and the plate 11. A coolant-introducing pipe 50 which is coupled to the delivery side of the impeller 49 of the main circulating pump 23 is communicated with the high-pressure plenum 48. A cylindrical partition 51 is laid around the outer periphery of the high-pressure plenum 48. An annular medium-pressure plenum 53 is defined between this cylindrical partition 51 and the cylinder 10. The high-pressure plenum 48 and the annular plenum 53 are communicated together by means of an orifice 52 provided in the cylindrical partition 51, as shown in FIG. 7. Furthermore, the annular plenum 53 and the medium-pressure plenum 47 are communicated by means of openings provided in the lower supporting plate 4. It can be said that the cylindrical partition 51 having the orifice 52 serves as a pressure reducing means for communicating between the medium-pressure plenum 47 and the high-pressure plenum 48. A partitioning wall 55 is provided in the reactor vessel 21. The partitioning wall 55 is installed on the periphery of the reactor core 1 and separates the inside space of the reactor vessel into the upper plenum 56 and the lower plenum 57. The low-temperature sodium in the lower plenum 57 in sucked by the main circulating pump 23, which increases the pressure of sodium, and is supplied to the inside of the high-pressure plenum 48 through the sodium-introducing pipe 50. The sodium in the high-pressure plenum 48 is supplied to the inside of each core fuel assembly 2A through the through hole 32 in the lower supporting plate 4. The sodium pressure P.sub.1 (refer to FIG. 5) is decreased to a predetermined pressure P.sub.2 (refer to FIG. 5) through the orifice 37. This sodium flows into the wrapper tube 35, is heated by heat generated by the fission of plutonium in the fuel pins, and is discharged from each core fuel assembly 2A to the upper plenum 56 in a high-temperature state. The high-temperature sodium discharged into the upper plenum 56 is introduced into the intermediate heat exchanger 22, where it is subjected to heat exchange with secondary sodium which therefore drops the temperature of the sodium from the upper plenum. The sodium whose temperature has dropped in the intermediate heat exchanger 22 is delivered again to the lower plenum 57 through the lower nozzles 58. In this embodiment, since the entrance nozzles of the fuel assemblies do not extend through the high-pressure plenum 48, unlike the case of the conventional example, the radial pressure loss of the high-pressure plenum 48 is extremely small. For this reason, the pressures at the entrance of the core fuel assemblies 2A are made virtually equal, together and the distribution of the flow rate in each core fuel assembly 2A is made uniform. Additionally, it is possible to reduce the height of the high-pressure plenum 48. Furthermore, the capacity of the main circulating pump 23 can be reduced as the radial pressure loss of the high-pressure plenum 48 can be reduced. The orifice 37 serves to adjust the distribution of the flow rate in each core fuel assembly 2A. According to this embodiment, the seismic resistance improves since transverse holes such as the conventionally employed ones are not present in the entrance nozzle 29. Next, description will be made of the effect of preventing the floating of the core fuel assemblies in the present embodiment. According to this embodiment, as shown in FIG. 5, the high-pressure receiving area of the entrance of each core fuel assembly 2A is made small (to reduce the floating force), and the area of the portion to which is applied a downward force due to the pressure differential between the internal pressure P.sub.2 of the core fuel assembly 2A and the pressure P.sub.3 of the medium-pressure plenum 47 is made large. In other words, the weight of the core fuel assembly 2A is supported on the annular line 38 (a circle with a diameter d.sub.4) of the annular bead 33. The portion of contact (concentric with the through hole 32) between the lower end of the entrance nozzle 29 of the core fuel assembly 2A and the annular bead 33 serves as a sealing portion. The outside area of the contacting portion S.sub.1 [=.pi.(D.sub.3.sup.2 -d.sub.4.sup.2)/4] is greater than the inside area of the contacting portion S.sub.2 [=.pi.(D.sub.4.sup.2 -d.sub.3.sup.2)/4]. The area S.sub.2 is the aforementioned high-pressure receiving area, while the area S.sub.1 is the area to which the aforementioned pressure differential (P.sub.2 -P.sub.3) is applied. For this reason, the downward force becomes large, and therefore it is possible to completely prevent the floating of the core fuel assemblies 2A since this force acts together with the weight of each core fuel assembly 2A itself. The height L of the entrance nozzle 29 according to this embodiment is approximately one-third of the height of the conventional entrance nozzle 15. A part of the sodium in the high-pressure plenum 48 reaches the inside of the annular plenum 53 after its pressure is appreciably reduced (by about one-tenth) through the orifice 52. The pressure in the annular plenum 53 is only slightly higher than that of the upper plenum 56. The sodium in the annular plenum 53 is introduced via its entrance nozzle into the blanket fuel assembly 2B supported by the lower supporting plate 4, similarly as shown in FIG. 5, and, at the same time, is introduced into the medium-pressure plenum 47 via the opening 54. The amount of heat generated in the blanket fuel assembly 2B is appreciably small in comparison with that of the core fuel assembly 2A; therefore, it is necessary to supply the sodium into the blanket fuel assembly 2B after reducing its pressure, as described above, in view of the distribution of the flow rate. The sodium which has come up in each blanket fuel assembly 2B is delivered to the upper plenum 56. The pressure in the medium-pressure plenum 47 is virtually equal to the pressure in the annular plenum 53. For this reason, the annular plenum 53 may be said to be a medium-pressure plenum. The sodium which has flown into the medium-pressure plenum 47 passes the openings 44 of the sleeve 45 and the opening 42 of the entrance nozzle 41 and then flows into the control rod guide pipe 39. This sodium comes up while cooling the control rod 27 and is delivered into the upper plenum 56. Since a multiplicity of entrance nozzles are present in the medium-pressure plenum 47, the radial pressure loss of the medium-pressure plenum 47 is large. For this reason, it is not necessary to provide an orifice for reducing the pressure in the entrance nozzle 41 of the control rod guide pipe, and as well the entrance nozzle 41 can be made short. As explained above, the supply of the sodium for cooling the control rod 27 from the medium-pressure plenum 47 is also advantageous in terms of the efficiency of inserting the control rods 27. In other words, it is necessary to rapidly insert the control rods 27 into the reactor core 1 at the time of a scram. When each control rod 27 is inserted, the sodium inside the control rod guide pipe 39 is discharged out by the volume of the inserted portion of the control rod 27. This sodium is discharged into the upper plenum 56 or into the medium-pressure plenum 47 in a counterflow manner. The channel resistance brought about by this discharged sodium constitutes a resisting force at the time of insertion of the control rod, and moderates the acceleration of the insertion of the control rod 27. If the sodium for cooling each control rod 27 is supplied from the high-pressure plenum, the counter flow of the sodium at the time of inserting the control rod occurs only with difficulty and since the sodium is discharged only into the upper plenum 56, there is a danger of the resistance becoming greater at the time of insertion. Meanwhile, if the sodium for cooling the control rod 27 is supplied from the medium-pressure plenum 47, it becomes readily possible for the sodium in the control rod guide pipe 39 to easily flow backward into the medium-pressure plenum 47 at the time of inserting the control rod. Hence, the resistance at the time of inserting the control rod can be made small. For this reason, according to the present embodiment, the efficiency of rapidly inserting the control rods 27 at the time of a scram can be improved, thereby making it possible to stop the fast breeder reactor in an extremely short period of time. According to this embodiment, since the heights of the entrance nozzles including the core fuel assembly, the blanket fuel assembly, and the control rod guide pipe can be greatly reduced, the height from the upper end of the reactor core 1 to the level of the sodium in the upper plenum 56 can be shortened by a greater degree than was possible in the past and the height is set to a predetermined value so that the upper ends of these fuel assemblies will not protrude from the sodium at the time when the fuel assemblies are taken out of the reactor core 1 and are moved horizontally inside the upper plenum 56. If the upper end of the fuel assemblies protrude from the level of the sodium, the cooling of the fuel assemblies is hampered at the time when they are moved horizontally. Accordingly, the lower the entrance nozzle 29, the height of the fuel assembly can be made smaller by the same margin, and the height from the upper end of the reactor core to the level of the sodium in the upper plenum can be reduced, and if the height becomes low, it becomes possible to lower the height of the reactor vessel 21 by the same margin. The height of the reactor vessel 21 can be further lowered by an amount equivalent to the amount by which the height of the high-pressure plenum 48 becomes lower, as described above. Thus, according to the present embodiment, it is possible to make the reactor vessel 21 compact and reduce the quantity of sodium contained therein. Another embodiment of the core fuel assembly is shown in FIG. 8. The figure shows a lower end portion of the core fuel assembly. According to this embodiment, in place of the orifice 37 of the core fuel assembly shown in FIG. 5, a neutron shielding body 60 which forms a labyrinth-like resisting body on its side is disposed at the lower end portion of the wrapper tube 35. This invention is also applicable to a loop-type fast breeder reactor. According to the present invention, it is possible to lower the height of a reactor vessel, and the efficiency of inserting control rods at the time of an emergency can be improved. |
048851230 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS A first embodiment of the present invention is shown in FIGS. 1a, 1b, 1c, 2 and 3. The first embodiment is described below. The apparatus for handling the core constituent elements of the present invention is provided on a small rotating plug 4 at the position where a conventional control rod drive mechanism is provided, in place of the conventional control rod drive mechanism. The same number of apparatuses for handling the core constituent elements as that of the conventional control rod drive mechanism are provided at the same positions. The apparatus for handling the core constituent elements has the arrangement described below. A cylindrical frame 40 is vertically provided in the small rotating plug 4 of a fast breeder reactor, and a first elevating drive mechanism is provided in an upper portion of the frame. The first elevating drive mechanism comprises a horizontal frame 50 which is fixed to the cylindrical frame 40 at an upper position of the frame and is provided with a first motor 42, a gear 51 which is provided on the rotary drive shaft of the first motor 42, and gears 52 which each engage with the gear 51 and which are each provided on a screw shaft 53 that is rotatably supported by the frame 50. A lower screw portion of each of the screw shafts 53 is screwed into a first elevating frame 54 which is connected to a sleeve 57 by a side frame 54a. The first elevating frame 54 supports a second elevating drive mechanism that comprises a frame 60 which is provided on the first elevating frame 54 and has a second motor 61, a gear 62 which is provided on the rotating drive shaft of the second motor 61, and other gears 63 which each engage with the gear 62 and which are each provided on a screw shaft 64 which has an upper portion rotatably supported by the frame 60 and a lower screw portion screwed into a second elevating frame 65. The second elevating frame 65 comprises a top frame 66, a cylindrical side frame 67 which is connected to the top frame 66, a frame 55 which is provided at the lower end of the side frame, bellows 70 which connect the frame 55 to an operational shaft 69, a gripper support frame 59 which has an upper end fixed to the frame 55, and bellows 58 which are connected to the frame 55 and one sleeve 56 of the sleeves 56, 57. The second elevating frame 65 supports a third elevating drive mechanism which comprises a frame 71 which is provided on a top plate 66 and has a third motor 72, a gear 73 which is provided on the rotary drive shaft of the third motor 72, and other gears which each engage with the gear 73 and which are each provided on a screw shaft 75 which has an upper portion rotatably supported by a frame 71 and a lower screw portion screwed into an operational shaft support frame 76. The upper end of the operational shaft 69 is fixed to the operational shaft support frame 76. The side frame 67 is, as shown in FIG. 1c, divided into an upper frame 77 and a lower frame 78 which are connected to each other by a magnetic link mechanism 79. This magnetic link mechanism 79 comprises an electromagnet 80 supported by the frame 77, a cylindrical body frame 81 attracted by the electromagnet 80, and links 82, 83 connecting the body frame 81 to the magnet side. Each of the links 82 engages with a projection of the frame 78. The operational shaft 69 comprises an upper operational shaft 84 and a lower operational shaft 85 which are connected to each other by the lower end of the upper shaft being tightly inserted into the upper end of the lower shaft at an intermediate position of the operational shaft 69. As shown in FIG. 2, the gripper support frame 59 has a lower end of an external diameter that can be inserted into a handling head 16 of each of control rod assemblies 13, and fuel assemblies 11, 12, but the upper portion thereof has a large external diameter which cannot be inserted into the handling head 16. As shown in FIG. 2, three grippers 87 are provided on shafts 86 at equal angular intervals within the gripper support frame 59, and are rotatable around the shafts 86 so that the state shown in FIG. 2 can be changed to the state shown in FIG. 3 or vice versa. Each of the grippers 87 has a shape in which projections 88, 89 projecting inwards are provided in the upper end and an intermediate portion of the gripper 87, respectively, an upper hook 90 projects outwards, and a lower hook 91 projects inwards. Since the hooks 90, 91 are separated from each other in the longitudinal direction but are not much separated from each other in the lateral direction, the grippers 87 can be inserted into the narrow handling head 16 even when the hooks 90, 91 are projecting outward and inward, respectively. A handling head 19 provided at the upper end of a control rod 18 contained in each of the control rod assemblies 13 has a shape having a projection which projects outward in the lateral direction and so can be supported by the lower hooks 91. The handling head 16 of each of the control rod assemblies 13 and the fuel assemblies 11, 12 has a shape having a projection which projects inward in the lateral direction and so can be supported by the upper hooks 90. The lower end of the operational shaft 69 is provided with an operational head 92 which comprises an upper portion 93, a lower portion 94, and an intermediate portion 95 having a greater diameter than those of the upper and lower portions. A description will now be made of the operation of the first embodiment as configured above. When the first motor 42 is driven, the torque of the motor 42 is transmitted to each of the gears so as to rotate the screw shafts. When the screw shafts are rotated, the first elevating frame 54 and the structural portions below the frame 54 can be moved either upward or downward. In a state wherein the first elevating frame 54 has been moved downward, the lower end of the gripper support frame 59 is inserted in the handling head 16. When the third motor 76 is then driven, each of the gears 73, 74 is rotated so as to rotate the screw shafts 75. When the frame 71 is moved downward by the rotation of the screw shafts 75, the operational shaft 69 moves downward, with the bellows 70 extending so that the intermediate portion 95 of the operational head 92 abuts against the projections 88. Consequently, each of the grippers 87 rotates around its shaft 86 so as to close the lower ends of the grippers 87. When the grippers 87 are closed, the handling head 19 of the control rod 18 is supported by the hooks 90 of the grippers 87 so as to be gripped thereby. When the second motor 61 is then driven so as to rotate each of the gears 62, 63, the screw shafts 64 are rotated so that the second elevating frame 65 is moved either upward or downward. In the state wherein the handling head 19 of the control rod 18 is gripped by the grippers 87, the longitudinal movement of the second elevating frame 65 causes the control rod 18 to move the same amount as the longitudinal movement in the same direction. When the control rod 18 is moved downward so as to be deeply inserted in the core of the reactor, the output of the reactor is decreased; while, when the control rod is shallowly inserted into the core, the output is correspondingly increased. In this way, the output of the reactor is controlled. If the output of the reactor increases abnormally, the temperature of the coolant of the reactor rises above the normal temperature. In this case, the function of the electromagnet 80 of the magnet link mechanism 79 is shut off so that the body frame 81 is moved downward, as shown by the dot-dot-dash lines in FIG. 1c, and the links 82, 83 which were placed at certain angles are aligned along an oblique line. As a result, the engagement between the link 82 and the frame 78 is removed so that the frame 78 is moved downward until it hits the upper end of the lower operational shaft 85. The impact force produced by this collision causes the lower operational shaft 85 to be separated from the upper operational shaft 84 and moved downward together with the frame 78. Consequently, the control rod 18 is also moved downward and is more deeply inserted into the core of the reactor at high speed, regardless of the motors. The output of the reactor is therefore decreased so that a safe state of the reactor is obtained. The control rod assemblies 13 and the fuel assemblies 11, 12, which are the core constituent elements, are handled in the manner described below. The first elevating frame 54 is moved downward by driving the first motor 42 so that the lower end of the gripper support frame 59 is inserted shallowly into each control rod assembly, as shown in FIG. 3. Then the operational shaft 69 is moved downward by driving the third motor 72 so that the upper portion 93 and the lower portion 94 of the operational head 92 are brought into contact with the projections 88 and 89 of the grippers, respectively, and the lower ends of the grippers are opened, as shown in FIG. 3. When the grippers are opened, the hooks 90 and the handling head 16 are engaged with each other. When the first elevating frame 54 is then moved upward by driving the first motor 42, the second elevating frame 65, the operational shaft 69, and the gripper support frame 50 are also moved upward by the same amount at the same time. Thus the grippers 87 are moved upward while they are still open, and the hooks 90 engage with the handling head 16 so that each of the control rod assemblies 13 and the various fuel assemblies 11, 12 can be pulled up out of the core of the reactor. Subsequently, each of the pulled-up control rod assemblies 13 and the fuel assemblies 11, 12 is placed above a pocket provided in the periphery of the core by horizontally rotating the large rotating plug and the small rotating plug 4, and is then made to fall into the pocket by driving the first motor in the reverse direction. The operational shaft is then moved upward by driving the third motor 72 so that the grippers are closed. The gripper support frame 59 is moved upward by driving the first motor 42 or the second motor 61 while the grippers are closed so that the grippers 87 are separated from the handling head 16. Each of new control rod assemblies 13 and various fuel assemblies 11, 12 placed in the pocket can be gripped by the grippers 87 and transferred into the core by an operation which is the reverse of that of the transfer from the reactor core to the pocket. Therefore, in accordance with the first embodiment, it is possible to control the output of the reactor by controlling the position in the core at which each control rod 18 is inserted, and to cause the emergency shutdown of the output of the reactor by inserting the control rods deeply into the core by adding the weight of the structure above each control rod 18 thereto, regardless of the motors which are separated by the magnetic link mechanism 79 in which the function of the magnet might be shut off in an emergency, facilitating the work of exchanging the core constituent elements. Since the control rods and the fuel assemblies, which are core constituent elements, can be handled and exchanged in the space where a conventional control rod drive mechanism is provided, the space required for a conventional fuel exchanger can be removed so that the diameter of the small rotating plug 4 and the diameter of the large rotating plug surrounding the small rotating plug 4 can be greatly reduced. Therefore, the size of the reactor vessel can be reduced and the amount of materials used in the reactor structure can also be reduced, so that the reactor structure can be made more economical. FIG. 4 shows a second embodiment of the present invention. Since in the second embodiment only the first elevating drive mechanism is changed from the first embodiment, only the changed part is described below. Racks 45 are vertically fixed to an upper wall of the frame 40, and pinions 43 which each engage with the racks 45 are rotatably provided on a frame 37. The first motor 42 is provided on the frame 37 which supports the first elevating frame 54, and one of the pinions 43 directly engages with a drive gear 97 which is rotated by the first motor 42 and is engaged with the other pinions 43 through an idle gear 98. In this structure, since each of the pinions 43 is rotated by driving the first motor 42 while the pinions 43 are engaged with the corresponding racks 45, the first frame 54 can be moved upward. Since the other parts are the same as those of the first embodiment, the second embodiment can achieve the same functional effect as the first embodiment. The third embodiment shown in FIG. 5 is also an embodiment in which only the first elevating drive mechanism is changed, but the other parts are the same as those in the first embodiment. Therefore, only the changed part is described below. The first motor 42 is provided on the frame 50 fixed to the frame 40, and one end of a rope 100 is fixed to the lower surface of the frame 50 by means of a rope clamp 99. An intermediate portion of the rope 100 is passed through a pulley 46 provided on the first elevating frame 54 and through a pulley 48 provided on the lower surface of the frame 50, and the other end of the rope 100 is fixed to a balance weight 101. A gear 102 rotated by the first motor 42 is engaged with a gear 103 having a shaft which is common with the pulley 48. Therefore, when the first motor 42 is driven, each of the gears 102, 103 is rotated so as to rotate the pulley 48. The first elevating frame 54 can be thus moved upward. Since the other parts are the same as those in the first embodiment, the third embodiment can achieve the same functional effect as the first embodiment. In accordance with the present invention, it is possible to remove the space required for a conventional fuel exchanger and reduce the size of a reactor. |
summary | ||
abstract | When an energy of a particle beam to be emitted from an accelerator is set for every slice group including two or more adjacent slices and an attenuation amount is set for each slice in the slice group, the energy to be emitted from the accelerator is set, for every slice group, higher than an energy corresponding to the slice at a deepest location in that slice group so that a transmissive plate has a predetermined thickness for the slice at the deepest location; and, with respect to a thickness of the transmissive plate to be set for every slice group, the thickness set for the slice group at a deep location is larger than or equal to the thickness set for the slice group at a shallow location, and the thickness set for the slice group at a deepest location is thicker than the thickness set for the slice group at a shallowest location. |
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051695936 | summary | BACKGROUND OF THE INVENTION The present invention relates to nuclear reactors and, more particularly, to tools for removing and installing control rod drives for commercial power nuclear reactors. A boiling-water nuclear reactor employs a plurality of fuel rods containing a nuclear fuel within a reactor vessel. The reactor vessel is filled with water to a level at least sufficient to cover the fuel rods. Fission in the fuel rods releases heat that boils the water surrounding them. This steam is used, either directly, or through an intermediate heat exchanger, to perform a useful function such as, for example, driving an electric turbine-generator. The intensity of the nuclear reaction in a nuclear reactor is controlled, in part, by moving control rods between fuel rods. The control rods absorb neutrons, thereby controlling the intensity of the nuclear reaction, and the rate at which steam is produced. The control rods are controlled by control rod drives inserted through the bottom of the reactor vessel. Control rod drives occasionally require maintenance or replacement. This has presented a problem because of the structure of the control rod drives and the working environment in which they must be handled. A typical control rod drive is about 16 feet long and weighs about 450 pounds. It is thus an awkward device that requires substantial mechanical handling assistance to install and remove. In addition, the sub-pile room below the reactor vessel typically has a headroom between the floor and the bottom of the reactor vessel of about 18 feet. This leaves little maneuvering room for lowering the control rod drive, rotating it into a horizontal position, and moving it out of the sub-pile room. Also, numerous fragile instrumentation cables hang down from the bottom of the reactor vessel. Such instrumentation cables can be damaged by contact with a control rod drive. If an instrumentation cable is damaged, the rules governing operation of a nuclear reactor require that work must stop until the damaged instrumentation cable is repaired. A further problem arises because the sub-pile room below a nuclear reactor is a high-radiation area. It is thus desirable to limit the amount of time that workers spend in that area. The following publications relate to devices which are used to lower and rotate a control rod drive in the sub-pile room. All of these publications are in Japanese, and full translations are not available. A translation of claim 1 is available and is provided for the use of the Patent and Trademark Office: Japanese Patent Publication No. 60-48715 PA0 Japanese Patent Publication No. SHO-60-49277 PA0 Japanese Patent Publication No. SHO-61-31839 PA0 Japanese Patent Publication No. 58-32359 PA0 Japanese Patent Publication No. SHO-61-36636 PA0 Japanese Patent Publication No. 61-42838 PA0 Japanese Patent Publication No. SHO-61-42839 PA0 Japanese Patent Publication No. SHO-61-36635 PA0 Japanese Patent Publication No. SHO-61-33158 PA0 Japanese Patent Publication No. SHO-61-25116 PA0 Japanese Patent Publication No. SHO-61-13198 PA0 Japanese Patent Publication No. SHO-57-39398 PA0 Japanese Patent Publication No. 57-49833 PA0 Japanese Patent Publication No. SHO-58-27880 PA0 Japanese Patent Publication No. 59-31034 PA0 Japanese Patent Publication No. SHO-60-35035 PA0 Japanese Patent Publication No. SHO-60-35036 PA0 Japanese Patent Publication No. 60-37439 PA0 Japanese Patent Publication No. SHO-60-46676 The length of the above list is regretted. However, the spirit of full disclosure requires the inclusion of each reference of which the applicants are aware. Also, as the seal between a control rod drive and the reactor vessel is broken during removal, a small amount of residual water spills from in the reactor vessel. Usually, the spilling water, which is contaminated with radioactivity, falls upon a worker in the process of removing the control rod drive. Although workers wear protective clothing and breathing apparatus in this area, it is considered undesirable to permit residual water to fall upon them. Japanese Utility Model Application Publication No. 57-49834, and Japanese Patent Publication Nos. SHO-58-15759 and SHO-53-18676 disclose water drain apparatus for use with control rod drives. Bolts securing a control rod drive are highly torqued during installation. Due to the cramped conditions in the sub-pile room, it is difficult to maneuver suitable tools into place to detorque these bolts to enable their removal. Japanese Patent Publication Nos. SHO-61-22274 and SHO-61-22275 disclose tools designed to remove such bolts. OBJECTS AND SUMMARY OF THE INVENTION It is an object of the invention to provide tools for handling control rod drives that overcome the drawbacks of the prior art. It is a further object of the invention to provide a handling tool for a control rod drive that permits positive control of the control rod during all stages of the removal process. It is a still further object of the invention to provide a handling tool for a control rod drive that reduces the likelihood of damaging instrumentation cables below a reactor vessel. It is a still further object of the invention to provide a handling tool for a control rod drive that reduces the time required for removing and installing a control rod drive. Briefly stated, the present invention provides a low-headroom tower that is pivotably mounted to a trunnion cart. The trunnion cart runs on rails in a slot in a work platform located in the sub-pile room of a reactor containment. An elevator in the tower raises an extension piece into contact with the bottom of a control rod drive. A detorquing guide is rotationally positioned to coincide with bolts holding the control rod drive in place. The elevator places an upward force on the control rod drive during detorquing of the bolts. This provides reaction torque to aid in bolt loosening and prevents leakage of contaminated effluent past the seal. A detorquing tool is fitted into the detorquing guide and is spring loaded to engage a selected bolt securing the control rod drive. An indexing device provides alignment for the detorquing tool with each succeeding bolt. The elevator is lowered until the bottom end of the control rod drive enters the tower. The load is transferred from the extension piece directly to the elevator. Lowering continues until the top end of the control rod drive emerges from the reactor vessel. A winch pivots the tower to the horizontal position about the trunnion cart, and rear wheels are engaged with the rails to permit rolling horizontal movement of the tower. An effluent container clamps around the control rod drive to channel away contaminated water that passes through the broken seal as the control rod drive experiences its first movement. A two-piece radiation shield pig is preset onto guide rods to clamp quickly onto the top end of the control rod drive. According to an embodiment of the invention, there is provided apparatus for handling a control rod drive for a nuclear reactor, comprising: a tower positionable below the nuclear reactor, means for lowering and raising the control rod drive a substantial distance within the tower, and means for rotating the tower, containing the control rod drive, between a horizontal and a vertical position, whereby transfer of the control rod drive is enabled. According to a feature of the invention, there is provided a method for handling a control rod drive for a nuclear reactor, comprising: positioning a tower below the nuclear reactor, lowering and raising the control rod drive a substantial distance within the tower, and rotating the tower, between a horizontal and a vertical position, whereby transfer of the control rod drive is enabled. According to a further feature of the invention, there is provided a method for removing a control rod drive from a nuclear reactor, comprising: positioning a tower below the control rod drive, engaging an upper end of an extension piece with the control rod drive, lowering the extension piece and the control rod drive a first portion of a distance required to clear the control rod drive from the nuclear reactor, removing the extension piece, continuing lowering the control rod drive a remainder of the distance until the control rod drive is clear of the nuclear reactor, and rotating the tower, with the control rod drive therein, to a horizontal position, whereby horizontal displacement of the control rod drive is enabled. According to a still further feature of the invention, there is provided a method for installing a control rod drive in a nuclear reactor, comprising: rolling a horizontal tower, containing the control rod drive, into position below the nuclear reactor, rotating the tower, and the control rod drive, into a substantially vertical position wherein a top end of the control rod drive is generally aligned with a predetermined point on a bottom of the nuclear reactor, raising the control rod drive a first portion of a distance required to install it in the nuclear reactor, transferring a load of the control rod drive to an extension piece, and continuing raising the control rod drive a remainder of a distance required to install it in the nuclear reactor. According to another feature of the invention, there is provided apparatus for removing a control rod drive from a nuclear reactor, comprising: a tower, means for positioning the tower below the control rod drive, an extension piece, means for engaging an upper end of the extension piece with the control rod drive, means for lowering the extension piece and the control rod drive a first portion of a distance required to clear the control rod drive from the nuclear reactor, means for removing the extension piece, means for continuing to lower the control rod drive a remainder of the distance until the control rod drive is clear of the nuclear reactor, and means for rotating the tower, with the control rod drive therein, to a horizontal position, whereby horizontal displacement of the control rod drive is enabled. According to still another feature of the invention, there is provided a torque breaker for breaking torque of a plurality of bolts securing a control rod drive of a nuclear reactor, the bolts being disposed in a first pattern, comprising: an extension piece, means for engaging the extension piece with a bottom of the control rod drive, a torque breaker tool, engagement means at a first end of the torque breaker tool, the engagement means being effective for rotationally engaging one of the plurality of bolts, an indexing guide affixed to the extension piece, support means at a second end of the torque breaker tool, pivoting means at a second end of the torque breaker tool, means in the indexing guide for pivotably engaging the pivoting means, the indexing guide including means for indexing to a plurality of predetermined positions about a circle, the plurality of predetermined positions being of the same number as the plurality of bolts, means for permitting rotation of the indexing guide to an angular position providing vertical alignment of one of the plurality of positions with one of the plurality of bolts, means for maintaining a fixed relationship between the indexing guide relative to the plurality of bolts, and means for permitting engagement of the engagement means with successive ones of the plurality of bolts, whereby torque of the plurality of bolts is broken. According to a still further feature of the invention, there is provided a torque breaker for breaking a torque of a plurality of bolts in a control rod drive, the bolts being disposed in a predetermined pattern, comprising: an extension piece, means for engaging the extension piece with a bottom of the control rod drive, a torque breaker tool, an indexing guide affixed to the extension piece, the indexing guide defining a plurality of positions corresponding to the predetermined pattern, means for aligning the indexing guide in an aligned position wherein one of the plurality of positions is aligned with one of the bolts, whereby all of the plurality of positions are aligned with corresponding bolt positions, means for locking the indexing guide in the aligned position, the indexing guide including means for retaining a bottom end of the torque breaker tool at any selectable one of the plurality of positions, an engaging portion at a top end of the torque breaker tool, the engaging portion including means for engaging an aligned one of the bolts, means for exerting torque on the torque breaker tool, whereby the one of the bolts is loosened, and means for indexing the torque breaker tool to a next one of the plurality of positions, whereby a next one of the bolts may be loosened. According to a still further feature of the invention, there is provided an effluent container for catching a burst of effluent from a nuclear reactor when a control rod drive is removed therefrom: a rod, means for moving the rod into forcible contact with a bottom of the control rod drive, the forcible contact being effective for avoiding substantial leakage of the effluent from the control rod drive, first and second halves of a water container, each of the first and second halves including a semi-cylindrical sidewall and a semi-circular bottom, each of the bottoms including a semi-circular cutout, the semi-circular cutouts being fitted together to form a circular hole generally conforming to a peripheral surface of the rod, means for conducting a liquid from the liquid container, the liquid container being fittable over a bottom of the control rod drive including a location from which effluent leakage is expected, a clamp cylinder fittable over the liquid container, the clamp cylinder being effective for holding the first and second halves of the liquid container together, and means on the clamp cylinder for permitting retention of the clamp cylinder while the liquid container is slid downward therethrough. According to a still further feature of the invention, there is provided a radiation shield pig assembly for shielding a filter end of a control rod drive as it exits a nuclear reactor, comprising: first and second guide rods affixed below the nuclear reactor adjacent opposed sides of the control rod drive, a first hanger assembly, first means for temporarily affixing the first hanger assembly on the first guide rod, a second hanger assembly, second means for temporarily affixing the second hanger assembly on the second guide rod, a first semi-cylindrical half shield, first quick-release means for affixing the first semi-cylindrical half shield to the first hanger assembly, a second semi-cylindrical half shield, second quick-release means for affixing the second semi-cylindrical half shield to the second hanger assembly, means for clamping abutting edges of the first and second semi-cylindrical half shields to form a cylindrical radiation shield, means for clamping the cylindrical radiation shield to the control rod drive, and means for releasing the first and second semi-cylindrical half shields from the first and second hanger assemblies, whereby the cylindrical radiation shield may remain on the control rod drive during movement thereof. According to a still further feature of the invention, there is provided a method for shielding an end of a control rod drive of a nuclear reactor, comprising: affixing first and second guide rods below the nuclear reactor adjacent opposed sides of the control rod drive, temporarily affixing a first hanger assembly on the first guide rod, temporarily affixing a second hanger assembly on the second guide rod, affixing a first semi-cylindrical half shield to the first hanger assembly, affixing a second semi-cylindrical half shield to the second hanger assembly, clamping together abutting edges of the first and second semi-cylindrical half shields to form a cylindrical radiation shield, clamping the cylindrical radiation shield to the control rod drive, and releasing the first and second semi-cylindrical half shields from the first and second hanger assemblies, whereby the cylindrical radiation shield may remain on the control rod drive during movement thereof. The above, and other objects, features and advantages of the present invention will become apparent from the following description read in conjunction with the accompanying drawings, in which like reference numerals designate the same elements. |
claims | 1. A method of communicating a capability rating for a computer product, the method comprising:assigning a capability rating to the computer product, wherein the assigning comprises associating a standard presentation of the capability rating with the computer product;determining the capability rating by a capability tool, executed by a processor, that determines the capability rating by comparing a set of computer system features obtained by the inventory module, performance results obtained by the performance testing module for the set of features, and rating level requirements that comprise a required set of features and required performance criteria;wherein the standard presentation of the capability rating comprises one of several integer numbers representing the capability rating, such that a higher integer number represents a higher capability rating; andproviding information on how the computer product exceeds the capability rating. 2. The method of claim 1 wherein the associating a standard presentation of the capability rating is performed under a licensing agreement. 3. The method of claim 1 wherein the standard presentation of the capability rating comprises a minimum capability rating and a recommended capability rating. 4. The method of claim 1 wherein the computer product is a computer hardware product. 5. The method of claim 1 wherein the computer product is a computer software product. 6. The method of claim 1 wherein the capability rating level is virtually identical to a software capability rating presentation such that a user can match the capability of software rated with the software capability rating presentation with the capability of the computer product. 7. The method of claim 1 wherein the capability rating comprises a minimum rating level, a recommended rating level and a best experience rating level. 8. The method of claim 1 wherein the capability rating is updated by a computer-executed capability tool associated with the computer product when the computer product is upgraded. 9. The method of claim 1 wherein the capability rating is determined as the capability rating of a component within the computer product with the lowest capability rating. 10. The method of claim 1 further comprising providing information on components required to upgrade the system to a higher capability rating. 11. A computer readable storage medium encoded with computer executable instructions for a software system for rating a computer system's ability to run software applications, wherein the rating facilitates matching of software application requirements with the computer system's capabilities, the software system comprising:an inventory module;a performance testing module; andan inventory and performance evaluator module;wherein the inventory and performance evaluator module compares:a set of computer system features obtained by the inventory module;performance results obtained by the performance testing module for the set of computer system features; andrating level requirements, wherein the rating level requirements comprise a required set of features and required performance criteria; andwherein the inventory and performance evaluator module determines a rating level for the computer system based on the comparison. 12. The computer readable storage medium of claim 11 further comprising a graphical user interface for displaying rating information. 13. The computer readable storage medium of claim 12 wherein the displayed rating information includes rating level requirements. 14. The computer readable storage medium of claim 12 wherein the displayed rating information includes rating levels of individual components in the computer system. 15. The computer readable storage medium of claim 12 wherein the displayed rating information includes a comparison of the performance results and the rating level requirements. 16. The computer readable storage medium of claim 12 wherein the displayed rating information includes recommended computer system upgrade information. |
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050540416 | abstract | An x-ray collimator for collimating an x-ray beam is constructed of a rotatable mandrel with a series of longitudinal slots of varying widths. The width of the collimated beam may be controlled by rotating the mandrel so that the correct slot lines up with the uncollimated x-ray beam. The angle of the beam may also be corrected by smaller angular rotations of the mandrel to offset the exit aperture of the slot. The entrance aperture of each slot is larger than the exit aperture so that such centerline adjustments do not affect the x-ray fan beam width. A very low backlash brake holds the mandrel against perturbing torques when collimator is in position. The brake includes a friction element and a means of reducing the torque of the positioning motor to reduce the effect of such perturbing torques. |
summary | ||
claims | 1. An ion implanter comprising: an ion source for generating an ion beam; at least one magnet disposed in the path of the ion beam for deflecting ions in the ion beam, said at least one magnet comprising first and second polepieces spaced apart to define a magnet gap through which the ion beam is transported; an electron source disposed on or in proximity to at least one of said polepieces for producing low energy electrons in the magnet gap; and a target site downstream of said at least one magnet for supporting a target for ion implantation, wherein the ion beam is delivered to said target site. 2. An ion implanter as defined in claim 1 , wherein the target is a semiconductor wafer. claim 1 3. An ion implanter as defined in claim 1 , wherein said at least one magnet comprises one magnet. claim 1 4. An ion implanter as defined in claim 1 , wherein said at lest one magnet comprises a plurality of magnets. claim 1 5. An ion implanter as defined in claim 1 , wherein the ion beam comprises a ribbon ion beam having a ribbon beam width and wherein said electron source produces low energy electrons across the ribbon beam width. claim 1 6. An ion implanter as defined in claim 1 , wherein the ion beam is scanned so as to produce an effective scan width and wherein said electron source produces low energy electrons across the scan width. claim 1 7. An ion implanter as defined in claim 1 , wherein said electron source comprises one or more linear electron sources disposed perpendicular to a direction of transport of the ion beam. claim 1 8. An ion implanter as defined in claim 1 , wherein said electron source comprises a one-dimensional array of electron sources. claim 1 9. An ion implanter as defined in claim 1 , wherein said electron source comprises a two-dimensional array of electron sources. claim 1 10. An ion implanter as defined in claim 1 , wherein said electron source comprises an area electron source. claim 1 11. An ion implanter as defined in claim 1 , wherein said electron source comprises an array of field emitters mounted to at least one of said polepieces and facing the magnet gap. claim 1 12. An ion implanter as defined in claim 1 , wherein said electron source comprises one or more electron-emitting wires disposed in proximity to at least one of said polepieces and perpendicular to a direction of transport of the ion beam. claim 1 13. An ion implanter as defined in claim 12 , wherein said magnet further comprises a polepiece liner and wherein said one or more electron-emitting wires are recessed in the polepiece liner. claim 12 14. An ion implanter as defined in claim 13 , wherein said one or more electron-emitting wires comprise tungsten wires and wherein said polepiece liner comprises graphite. claim 13 15. An ion implanter as defined in claim 12 , wherein said one or more electron-emitting wires are recessed in at least one of said polepieces. claim 12 16. An ion implanter as defined in claim 12 , wherein said electron source further comprises an insulator disposed behind each of said one or more electron-emitting wires for reflecting electrons toward the ion beam. claim 12 17. An ion implanter as defined in claim 12 , wherein said electron source further comprises an electrically isolated conductor disposed behind each of said one or more electron-emitting wires for reflecting electrons toward the ion beam. claim 12 18. An ion implanter as defined in claim 17 , further comprising a bias power supply connected to said conductor for controlling the efficiency of electron reflection. claim 17 19. An ion implanter as defined in claim 1 , wherein said electron source is mounted on at least one of said polepieces in the magnet gap. claim 1 20. An ion implanter as defined in claim 1 , wherein said electron source is located between said first and second polepieces. claim 1 21. An ion implanter as defined in claim 1 , wherein said electron source is recessed in at least one of said polepieces. claim 1 22. A method for transporting an ion beam through a magnet, comprising: directing the ion beam through a magnet gap between first and second polepieces of the magnet; and supplying low energy electrons to the ion beam being transported through the magnet gap between the first and second polepieces of the magnet by an electron source disposed on or in proximity to at least one of the first and second polepieces. 23. A method as defined in claim 22 , further comprising the step of transporting the ion beam from the magnet to a target at a target site. claim 22 24. A method as defined in claim 22 , wherein the step of supplying low energy electrons comprises supplying electrons from one or more electron sources disposed perpendicular to a direction or transport of the ion beam. claim 22 25. A method as defined in claim 22 , wherein the step of supplying low energy electrons comprises supplying electrons from one or more electron-emitting wires disposed in proximity to at least one of the polepieces. claim 22 26. A magnet assembly for operation with an ion beam, comprising: a magnet disposed in the path of the ion beam, said magnet comprising first and second polepieces spaced apart to define a magnet gap through which the ion beam is transported; and one or more electron sources disposed on or in proximity to at least one of said polepieces for producing low energy electrons in the magnet gap. |
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description | The present disclosure relates generally to method for sealing an opening extending radially from an outer circumferential surface to an inner circumferential surface of a tubular object in a nuclear power plant, more specifically for sealing an opening in a portion of a feedwater sparger in a nuclear pressure vessel. U.S. Pat. No. 5,839,192 discloses clamping the outside of a BWR sparger, U.S. Pat. No. 7,871,111 discloses repairing flawed welded joint in a core spray piping system, U.S. Pat. No. 4,573,628 discloses a method for tapping into tubing of a nuclear power station, U.S. Pat. No. 5,408,883 discloses cutting an elliptical hole in nuclear steam generator tubing, U.S. Pat. No. 6,456,682 discloses a BWR core spray sparger T-box attachment with clamp. A method for sealing an opening extending radially from an outer circumferential surface to an inner circumferential surface of a tubular object in a nuclear power plant includes inserting a stopper from outside of the outer circumferential surface through the opening into the tubular object; and actuating a fastener from the outside of the circumferential surface to force the stopper radially outward to seal the opening. A method for removing material from an inaccessible region of an enclosure in a nuclear power plant includes cutting through a wall of the enclosure to form an opening passing through the wall; removing the material through the opening; inserting a stopper into the enclosure through the opening from outside of the enclosure; and then actuating a fastener from the outside of the enclosure to force the stopper against the wall to seal the opening. A mechanical seal assembly for plugging an opening in a tubular object by contacting an inner circumferential surface of the tubular object includes a stopper configured for insertion into an interior of the tubular object for plugging the opening. The stopper includes a surface configured for matching the inner circumferential surface of the tubular object. The mechanical seal assembly also includes a fastener passing through a hole in the stopper such that the fastener is actuatable from outside of the tubular object to force the surface of the stopper against the inner circumferential surface of the tubular object. The present disclosure provides a mechanical seal assembly configured for sealing an opening in tubular object in a nuclear power plant. In particular, the mechanical sealing assembly can be used on a six inch curved pipe of a feedwater sparger of a boiling water reactor (BWR). The mechanical seal assembly can be used to seal a hole that has been cut into a tubular object to remove foreign material trapped in the tubular object. In particular, to remove foreign material from a tubular object in a nuclear power plant, in particular in a location that is inaccessible by Foreign Object Search and Retrieval (FOSAR) tools, a hole can be cut that extends from an outer circumferential surface to an inner circumferential surface of the tubular object. The foreign material can be removed from the tubular object through the hole, then the mechanical seal assembly can be used to seal the hole. The use of the such a technique to remove the foreign material and seal the hole may have a number of advantageous features: the mechanical seal assembly can include a precision machined seal plate mounted inside a sparger, a strongback mounted outside the sparger and two bolts with integrated crimp cups to hold it all together; the strongback is not welded to the pipe, but is interlocked between the external pipe surface surrounding the pipe hole, and the bolts of the seal plate associated with the edge of the seal plate can be blocked to the internal perimeter of the pipe hole, the seal plate can be designed to fit inside a curved pipe; the seal plate can be tuned to provide a specific flexure to maintain bolt pre-load over the life of the system; the bolts can held in place by an expanding crimp cup design; the EDM hole can be dimensioned to allow the seal plate to be introduced inside the pipe in a given position so the seal plate can be then moved in a sealing position where the edge of the seal plate is in contact with the internal perimeter of the pipe hole, the mechanical seal assembly can be designed to be delivered and installed remotely, underwater; in a BWR environment; the mechanical seal assembly can be a permanent solution, but it can be removed in the future if required; the mechanical seal assembly can provide a 100% leak proof seal or it can be designed to be less than a 100% leak proof seal, depending on what is acceptable per analysis of feedwater flow through the sparger; the mechanical seal assembly can be implemented with fuel in place in the vessel; and/or the mechanical seal assembly can maintain radiation exposure ALARA. FIG. 1 shows a perspective view of a portion of a BWR feedwater sparger assembly 10 with an opening 12 cut therein in accordance with an embodiment of the present invention. In one preferred embodiment, the opening 12 is cut via electrical discharge machining (EDM) apparatus, but other cutting techniques may also be used, for example using a machine tool, or a saw. As schematically shown in FIG. 1, BWR feedwater sparger assembly 10 includes an enclosure in the form of a curved tube 16, with a plurality of nozzles 18 protruding from a top 20a of an outer circumferential surface 20 of curved tube 16. Curved tube 16 is considered as having a curved shape because its center axis CA follows an arced path while extending longitudinally from a first end 22 to a second end 24, such that an interior edge 20b of outer circumferential surface 20—i.e., an edge facing a center axis of the pressure vessel—has a concave shape when extending axially with respect to center axis CA and an exterior edge 20c (FIGS. 3, 4) of outer circumferential surface 20—i.e., an edge facing away from the center axis of the pressure vessel—has a convex shape when extending axially with respect to center axis CA, as seen in the radially facing cross-sectional view of FIG. 3. Opening 12 is cut into an interior facing portion 20d of outer circumferential 20 that includes interior edge 20b by removing material from a curved wall 17 at interior facing portion 20d. Using EDM, opening 12 is cut without creating foreign material (FM). Curved wall 17 is cut through from outer circumferential surface 20 to an inner circumferential surface 40 (FIG. 3, 4) of tube 16 to form opening 12. Inner circumferential surface 40 at wall 17 has a convex shape when extending axially with respect to center axis CA, as seen in the radially facing cross-sectional view of FIG. 3, and a concave shape when extending circumferentially with respect to center axis CA, as seen in the axially facing cross-sectional view of FIG. 3. EDM may use electrical discharges to remove material from curved wall 17 of curved tube 16 to form an elliptical cut creating the elliptical opening 12. After opening 12 is cut into tube 16, a tool or a human operator's hand may be inserted through opening 12 into an interior 21 of tube 16 to remove the foreign object trapped inside of tube 16. As shown in FIG. 1, opening 12 includes a length L1 extending between longitudinal edges 12a, 12b and a width W1 extending between lateral edges 12c, 12d that is less than length L1. Length L1 defines a maximum perimeter dimension of opening 12 at outer circumferential surface 20 and width W1 defines a minimum perimeter dimension of opening 12 at outer circumferential surface 20. FIGS. 2 to 6 show views of a mechanical seal assembly 30 in accordance with a first embodiment of the present invention sealing the opening 12 cut into curved tube 16. Seal assembly 30 includes a stopper 32 formed as a stainless steel machined plate, at least one fastener in the form of two bolts 34 and a strongback 36 for supporting bolts 34. As shown in FIG. 6, which shows a perspective view of stopper 32 and bolts 34, stopper 32 has a length L2 extending between longitudinal edges 32a, 32b and a width W2 extending between lateral edges 32c, 32d that is less than length L2. Length L2 defines a maximum perimeter dimension of stopper 32 and width W2 defines a minimum perimeter dimension of stopper 32. Width W2 of stopper 32 is less than length L1 of opening 12 such that stopper 32 can be oriented widthwise and inserted through opening 12 into interior 21 of tube 16. Length L2 of stopper 32 is greater than length L1 of opening 12 and width W2 of stopper 32 is greater than width W1 of opening 12 such that stopper 32 can plug opening 12. Stopper 32 includes an outer ring 38 whose outer extent defines an outer perimeter 38a of stopper 32 having an elliptical shape. Outer ring 38 includes an exterior surface 38b for facing away from interior 21 of tube 16 that is configured for contacting an inner circumferential surface 40 of tube 16. Surface 38b has a curved shape that defines two convex surface portions 38c, 38d extending widthwise and two concave surface portions 38e, 38f extending lengthwise, which allow surface 38b to match the shape of the inner circumferential surface 40 of tube 16, due to the curved shape of tube 16 described above with respect to FIG. 1. A first convex surface portion 38c extends from lateral edge 32d past longitudinal edge 32a to lateral edge 32c to define a convex shape and a second convex surface portion 38d extends from lateral edge 32d past longitudinal edge 32b to lateral edge 32c to define a convex shape. A first concave surface portion 38e, which overlaps with portions 38c, 38d, extends from longitudinal edge 32a past lateral edge 32d to longitudinal edge 32b to define a concave shape and second concave surface portion 38f, which overlaps with portions 38c, 38d, extends from longitudinal edge 32a past lateral edge 32c to longitudinal edge 32b to define a concave shape. Stopper 32 also includes a central portion 42 inside of outer ring 38. Central portion 42 is also used as an aid in positioning stopper 32 in opening 12, by allowing a remote operator to know if the stopper 32 is installed to completely seal the opening 12, i.e., confirming that stopper 32 is not shifted too high or too low in the pipe. Central portion 42 joins an inner perimeter of outer ring 38 and protrudes away from outer ring 38 such that central portion 42 is thicker than outer ring 38 and extends partially into opening 12. Central portion 42 includes an exterior surface 42a for facing away from interior 21 of tube 16 that, similar to outer ring 38, has a concave shape extending in the lengthwise direction and a convex shape extending in the widthwise direction. Central portion 42 also has an elliptical shape perimeter that defines two longitudinal edges 42b, 42c and two lateral edges 42d, 42e. A handle 44 is fixed to exterior surface 42a for gripping by a tool or human operator. Stopper 32 includes threaded holes 46 extending therethrough configured for receiving threaded shafts 34a of bolts 34. On an interior side thereof, stopper 32 includes protrusions 48 forming a thickened portion of stopper 32 to extend the length of threaded holes 46. Strongback 36 is configured for contacting outer circumferential surface 20 of tube 16. More specifically, strongback 36 includes a plurality of contact surfaces 50 configured for contacting outer circumferential surface 20. Each contact surface 50 is formed at an end 52a, 52b, 52c, 52d of one of legs 54a, 54b of strongback 36. More specifically, strongback 36 includes two legs 54a, 54b, with a first leg 54a including two ends 52a, 52b for contacting surface 20 and a second leg 54b including two ends 52c, 52d for contacting surface 20. Strongback 36 also includes a central bar 56 connecting legs 54a, 54b. Central bar 56 includes two slots 58 formed therein, each for receiving one of bolts 34. Slots 58 are each surrounded by a flange 60 recessed below an exterior surface 56a of central bar 56. Flanges 60 form shoulders each configured for contacting an annular collar 34b of a head 34c of the respective one of bolts 34. A method of installing seal assembly 30 will now be described. First, stopper 32 is orientated such that the width of stopper 32 is aligned with the length of opening 12 and stopper 32 is passed through opening 12 such that lateral edges 32c, 32d of stopper 32 face longitudinal edges 12a, 12b of opening 12 as stopper 32 is being passed through opening 12. More specifically, one of the two longitudinal edges 32a, 32b of stopper is first passed through opening 12, then lateral edges 32c, 32d are passed through opening 12, and then lastly the other of the two longitudinal edges 32a, 32b is passed through opening 12. After stopper 32 is passed through opening 12 into interior 21 of tube 16, stopper 32 is rotated such that exterior surface 42a is facing opening 12, and then stopper 32 is pulled toward opening 12 so that exterior surface 38b of outer ring 38 contacts inner circumferential surface 40 of tube 16. As outer ring 38 contacts inner circumferential surface 40 of tube 16, central portion 42 penetrates at least partially into opening 12 such that longitudinal edges 42b, 42c of central portion 42 face and are directly adjacent to longitudinal edges 12a, 12b, respectively, of opening 12 and lateral edges 42d, 42e of central portion 42 face and are directly adjacent to lateral edges 12c, 12d, respectively, of opening 12. After stopper 32 is inserted into opening 12, heads 34c and parts of shafts 34a of bolts 34 protrude outside of tube 16. Strongback 36 is then slid onto bolts 34 such that shafts 34a of bolts 34 are received in slots 58. Then, the bolts 34 are actuated from the outside of outer circumferential surface 20 of tube 16 to force stopper 32 radially outward with respect to center axis CA to seal opening 12. More specifically, heads 34c of bolts 34 are torqued with a tool such that shafts 34a move further into threaded holes 46 and annular collars 34b of bolts 34 are forced into flanges 60. This torquing of bolts 34 into a tightened position tensions bolts 34 and pulls strongback 36 and stopper 32 closer together such that exterior surface 38b of outer ring 38 of stopper 32 is pressed tightly against inner circumferential surface 40 of tube 16 and surfaces 50 of strongback 36 are pressed tightly against outer circumferential surface 20 of tube 16. This tightening of bolts 34 causes stopper 32 to sealingly engage inner circumferential surface 40 of tube 16 to minimize leakage out of tube 16 during operation of sparger assembly 10. Locking cups (similar to locking cups 134f shown in FIGS. 9 and 10) are utilized to secure the bolts 34. The locking cups are co-axial with the head 34c of the bolts 34, and a crimping tool is used to crimp the locking cups around the head 34c, blocking bolt 34 in position. The method described with respect to FIGS. 2 to 6 can be accomplished by delivering the stopper 32, with bolts 34 preinstalled in holes 46, via a first long pole handling tool to sparger assembly 10 inside the pressure vessel and then delivering strongback 36 via a second long pole handling tool and holding the strongback 36 in place via the second long pole handling tool. Then, a third long pole handling tool with a right angle drive tool can be used to tighten the bolts 34 to apply a clamping load. Finally, the crimping tool can be used to crimp the locking cups. The long pole handling tools can be maneuvered from a refueling bridge or an auxiliary bridge located above the pressure vessel. FIGS. 7 to 11 show views of a mechanical seal assembly 130 in accordance with a second embodiment of the present invention sealing the opening 12 cut into curved tube 16. Seal assembly 130 includes a stopper 132 formed as curved plate and at least one fastener in the form of two bolts 134. Same as stopper 32, stopper 132 includes a length L2 extending between longitudinal edges 132a, 132b and a width W2 extending between a lateral edge 132c, 132d that is less than length L2, with length L2 defining a maximum perimeter dimension of stopper 132 and width W2 defines a minimum perimeter dimension of stopper 132. Width W2 of stopper 132 is less than length L1 of opening 12 such that stopper 132 can be oriented widthwise and inserted through opening 12 into interior 21 of tube 16. Length L2 of stopper 132 is greater than length L1 of opening 12 and width W2 of stopper 132 is greater than width W1 of opening 12 such that stopper 132 can plug opening 12. Stopper 132 is formed in substantially the same manner as stopper 32, with the differences from stopper 32 being the configuration of holes 146 for receiving bolts 134 in stopper 132, stopper 132 including an exterior support assembly 170 for bolts 134 and stopper 132 not having a handle 44. Accordingly, stopper 132 includes an outer ring 138 formed in the same manner as outer ring 38 and a central portion 142 formed in the same manner as central portion 42. Stopper 132 includes threaded holes 146 extending therethrough configured for receiving threaded shafts 134a of bolts 134. On an interior side thereof, stopper 132 includes protrusions 148 formed in the same manner as protrusions 48 to extend the length of threaded holes 146. Bolts 134 each include a foot 134d formed at a distal end thereof for contacting inner circumferential surface 40 of tube 16 directly opposite of opening 12. Each foot 134d extends past the outer diameter of the respective shaft 134a. Each foot 134d is includes a curved contact surface 134e such that the foot 134d can maximize contact between contact surface 134e of foot 134d and inner circumferential surface 40. Bolts 134 each include a head 134c formed at a proximal end thereof for being received in exterior support assembly 170 of stopper 132. Each head 134c includes an annular collar 134b and has a locking cup 134f non-rotatably attached thereto for securing the bolt 134 in place after installation. Exterior support assembly 170 includes a two receptacles 172 protruding from an exterior surface 142a of central portion 142 and a support bar 174 connecting receptacles 172. Each receptacle 172 is cylindrically shaped and includes an inner cylindrical surface 172a and an annular surface 172b defining a bore therein. When bolts 134 are installed in their tightened position in receptacles 172, heads 134c and locking cups 134f are received in receptacles 172 and annular collar 134b abuts annular surface 172b. Locking cups 134f are then forced radially outward with respect to a center axis of the respective bolt 134, about which the bolt 134 is rotated about during installation, by a crimping tool into form-fitting contact with cylindrical surface 172a such that bolts 134 are secured in place at are prevented from loosening. A method of installing seal assembly 130 will now be described. First, stopper 132 is orientated such that the width of stopper 132 is aligned with the length of opening 12 and stopper 132 is passed through opening 12 such that lateral edges 132c, 132d of stopper 132 face longitudinal edges 12a, 12b of opening 12 as stopper 132 is being passed through opening 12. More specifically, as shown in FIG. 11, one of the two longitudinal edges 132a, 132b of stopper 132 (edge 132a in FIG. 11) is first passed through opening 12, then lateral edges 132c, 132d are passed through opening 12, and then lastly the other of the two longitudinal edges 132a, 132b (edge 132b in FIG. 11) is passed through opening 12. After stopper 132 is passed through opening 12 into interior 21 of tube 16, stopper 132 is rotated such that exterior surface 142a is facing opening 12, and then stopper 32 is pulled toward opening 12 so that exterior surface 138b of outer ring 138 contacts inner circumferential surface 40 of tube 16. As outer ring 138 contacts inner circumferential surface 40 of tube 16, central portion 142 penetrates at least partially into opening 12 and exterior support assembly 170 extends out of opening 12 past outer circumferential surface 20. After stopper 132 is inserted into opening 12, heads 134c and parts of shafts 134a of bolts 134 protrude outside of tube 16. Bolts 134 are then actuated from the outside of outer circumferential surface 20 of tube 16 to force stopper 132 radially outward with respect to center axis CA to seal opening 12. More specifically, heads 134c of bolts 134 are torqued with a tool such that shafts 134a move further into threaded holes 146, feet 134d are forced into inner circumferential surface 40 and annular collars 134b of bolts 134 are forced into annular surfaces 172b. This torquing of bolts 134 into a tightened position compresses bolts 134 and pulls stopper 132 away from feet 134d and toward opening 12 such that exterior surface 138b of outer ring 138 of stopper 132 is pressed tightly against inner circumferential surface 40 of tube 16. This tightening of bolts 134 causes stopper 132 to sealingly engage inner circumferential surface 40 of tube 16 to minimize leakage out of tube 16 during operation of sparger assembly 10. Cups 134f can then be forced radially outward with respect to the center axis of the respective bolt 134 by the crimping tool into form-fitting contact with cylindrical surface 172a such that bolts 134 are secured in place and prevented from loosening. The method described with respect to FIGS. 7 to 11 can be accomplished by delivering the stopper 132, with bolts 134 preinstalled in holes 146, via a first long pole handling tool to sparger assembly 10 inside the pressure vessel and holding the stopper 132 in place via the first long pole handling tool. Then, a second long pole handling tool with a right angle drive tool can be used to tighten the bolts 134 to apply a clamping load. Finally, the crimping tool can be used to crimp the locking cups 134f. FIGS. 12a to 12c show views of a stopper 232 in accordance with another embodiment of the present invention. Stopper 232 is used in the same manner as stopper 132, with bolts 134, but includes a groove 200 formed therein receiving a gasket 202. Stopper 232 includes a central portion 242 formed in a similar manner as central portions 42, 142 and an outer ring 238 formed in a similar manner as outer rings 38, 138, with a main difference being that outer ring 238 is provided with a groove 200 formed in exterior surface 238b of outer ring 238. Groove 200 is continuous and completely surrounds central portion 242. Gasket 202 is provided in groove 200 and completely surrounds central portion 242. When stopper 232 is installed on tube 16, as shown in FIGS. 12a and 12b, gasket 202 contacts inner circumferential surface 40 of tube 16 and completely surrounds opening 12. It should be noted that stoppers 32, 132 and 332 (described below) can be provided with a groove and gasket in the same manner as groove 200 and gasket 202. FIG. 13 shows a view of a stopper 332 in accordance with another embodiment of the present invention. Stopper 332 is formed in substantially the same manner as stopper 32, except that stopper 332 includes a two-piece strongback 336. Strongback 336 includes a first piece 336a configured for contacting outer circumferential surface 20 of tube 16 and a second piece 336b for receiving heads 334c of bolts 334. More specifically, strongback piece 336a includes a plurality of contact surfaces 350 configured for contacting outer circumferential surface 20. Each contact surface 350 is formed at an end 352a, 352b, 352c, 352d of one of legs 354a, 354b of strongback piece 336a. Strongback piece 336a also includes a central bar 356 connecting legs 354a, 354b. Strongback piece 336b extends along central bar 356 from leg 354a to leg 354b and includes two slots 358 formed therein, each for receiving one of heads 334c of bolts 334. Strongback piece 336b includes a convex surface 370 that is received in a concave groove 372 in piece 336a. The two piece construction of strongback 336 optimizes stress distribution in the strongback piece 336b contacting the tube 16 as bolts 334 are tightened to allow precise control over the preload due to the asymmetrical nature of the strongback 336. In the preceding specification, the invention has been described with reference to specific exemplary embodiments and examples thereof. It will, however, be evident that various modifications and changes may be made thereto without departing from the broader spirit and scope of invention as set forth in the claims that follow. The specification and drawings are accordingly to be regarded in an illustrative manner rather than a restrictive sense. |
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048428060 | claims | 1. In a boiling water nuclear reactor having a reactor pressure vessel with a bottom, an internal recirculation pump mounted in said bottom of said vessel, said vessel being at least partially filled with a primary coolant, a reactor core disposed diagonally above said internal recirculation pump, and means for measuring the recirculation flow rate of said coolant, comprising: a flow passage resisting means disposed in a recirculating flow passage of the primary coolant in the reactor pressure vessel and adapted to cause, between an upstream side and a downstream side thereof, a pressure difference in the primary coolant while the flow of the primary coolant is maintained constant; means disposed at an upstream position and a downstream position relative to said flow passage resisting means for providing a first detecting means for detecting a value of said pressure difference; means operatively connected to said first detecting means for converting the detected pressure difference value into an electrical signal representing the pressure difference and transmitting the electrical signal; and means operatively connected to said transmitting means for computing the recirculating flow rate of the primary coolant in response to the electrical signal from said transmitting means; wherein said flow passage resisting means comprises a plurality of support legs which extend between the bottom of the reactor pressure vessel and a lower end part of a cylindrical shroud disposed within the reactor pressure vessel to partition the interior thereof and serving to support said lower end part of said shroud, and said differential pressure detecting means are disposed, respectively, at upstream and downstream sides of said shroud support legs which are interposed therebetween. an auxiliary measuring means for measuring a core flow rate of the primary coolant, said auxiliary measuring means comprising means located at upstream and downstream sides of a core support plate and adapted to provide a second detecting means to detect the pressure difference of the primary coolant therebetween; means operatively connected to said second detecting means and adapted to convert said detected pressure difference value into an electrical signal and transmitting the electrical signal; and means operatively connected to said transmitting means and operating to compute the core flow rate in response to the signal from said transmitting means. 2. The apparatus according to claim 1 further comprising: |
054141974 | abstract | A method of containing hazardous and toxic wastes includes the steps of irporating the dried waste, in a salt form, in melted polymer, such as asphalt, and forming the waste salt and asphalt blend into aggregate pellets. The pellets are coated with a powdered coating material that is compatible with a portland cement-based mortar or other cementitious material which is used. The coated particles are mixed with mortar to form a polymer-aggregate concrete and cast into wasteforms for storage or burial. If it is desirable to produce a waste form with a continuous layer of mortar on the exterior of the concrete monolith the mold can be placed on a turntable and spun, or otherwise exposed to a centrifugal force to force the mortar to the outside of the mold. Centrifugal separation is possible because the polymer-waste mixture typically has a specific gravity near 1.5 while that of the cementitious mixture is typically greater than 2.0. |
054854911 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention is directed to a diagnostic system for a rotating electrical apparatus, and more particularly to an online system for diagnosing the condition and maintenance requirements of an operational motor. 2. Background of Information Large motors require periodic preventive maintenance to ensure reliable and efficient performance over an operational lifetime. The requisite maintenance period is a function of a specific design of a particular motor. Reactor coolant pump (RCP) motors used in nuclear power plants, for example, are typically inspected annually, during refueling outages, and are disassembled for inspection or refurbishment every five to ten years. Significant time and field effort are required to disassemble such motors and perform the requisite inspections. A set of motors having an identical design are, nevertheless, different in terms of a variety of factors such as normal manufacturing tolerances, operating history and environment, and quality of maintenance. A pre-specified operating period, before motor inspection or maintenance, is an informed estimate of the appropriate operational schedule for a typical motor under recommended operating conditions. In some cases, the pre-specified operating period may be too long (e.g., where recommended motor operating parameters have been exceeded) and, hence, inefficient operation may result at the end of the period. In other cases, the prespecified operating period may be too short (e.g., where the motor has been relatively lightly loaded) and, therefore, unnecessary costs associated with outage time, manpower and material will be expended. Furthermore, where a RCP motor is located within a nuclear containment vessel, unnecessary human exposure to radiation would result. There is a need, therefore, for a system which accurately diagnoses the condition of an operational rotating electrical apparatus. There is a more particular need for such a system which accurately diagnoses the maintenance or inspection requirements of the operational rotating electrical apparatus. SUMMARY OF THE INVENTION These and other needs are satisfied by the invention which is directed to an online system for diagnosing operating conditions of a motor, in order to determine when motor maintenance is required. Motor sensors are provided which monitor various physical parameters (e.g., non-electrical or insulation-related conditions) and produce corresponding electrical signals. Signal converters transform the electrical signals to corresponding digital values. These values are collected by a processor which compares the values, or a trend of the values, with predetermined baseline values, or trends, associated with a newly manufactured or refurbished motor. The processor then makes recommendations for a motor maintenance interval, in order to provide optimum motor performance and availability at minimum cost and downtime. The motor maintenance interval is a specific time or, alternatively, a more general time, such as the time of the next scheduled refueling outage. In the case of a RCP motor within a nuclear containment vessel, an intermediate data storage device collects the digital values corresponding to the electrical signals and communicates the digital values to a processor which is remotely located (e.g., beyond a biological barrier, beyond the containment vessel, at an off-site location, etc.) from the RCP motor. |
description | This application is a Divisional Application of U.S. application Ser. No. 13/531,315, filed Jun. 22, 2012, which is a Bypass Continuation of PCT/EP2010/068476, which has an International Filing Date of Nov. 30, 2010, and which claims priority under 35 U.S.C. 119(a) to German Patent Application No. 10 2009 055 119.0, filed on Dec. 22, 2009. The entire contents of each of these related applications are hereby incorporated by reference into the disclosure of the present divisional application. The invention relates to production methods for mirror elements and to mirror elements comprising a reflective coating for the EUV wavelength range and a substrate. It is known that the density of many materials, especially materials containing silicon, changes under irradiation with high-energy radiation. That effect is referred to in the literature as “compaction”. For applications in extreme environments (reactors, outer space) in particular, studies were carried out and those effects were quantitatively determined a long time ago (see W. Primak, Nucl. Sci. Eng., 65, 141, 1978: “Radiation Behaviour of Vitreous Silica” and R. A. B. Devine, “Macroscopic and Microscopic Effects of Radiation in Amorphous SiO2”, Nuclear Instruments and Methods in Physics Research B 91 (1994), 378-390). It has been found that the change in volume or density in silicon dioxide typically attains, after sufficiently long irradiation, a saturation value on the order of about 2%-3% within the penetration depth reached by the radiation. The penetration depths of the high-energy types of radiation considered therein were typically in a range of from 0.5 μm to about 10 μm or more. Comparable effects are also known in microlithography, especially for the VUV wavelength range; however, owing to the relatively low level of interaction of the VUV light with the optical material used, especially quartz glass, the changes in volume in that case are as a rule in the ppm range and therefore, typically, no saturation value is attained, the optical materials used in that case being completely penetrated by the radiation. A method is known from U.S. Pat. No. 6,205,818 B1 by which quartz glass (SiO2) is to be made insensitive to compacting caused by long-term irradiation with UV laser radiation. The method provides for the quartz glass material to be pre-compacted by being exposed to high-energy radiation or by being pre-treated by hot isostatic pressing (HIP). The high-energy radiation is said to make a compaction of between about 10 ppm and 100 ppm possible, while a change in volume of the entire quartz glass body of from about 0.1% to about 3% is said to be achievable by hot isostatic pressing. Since microlithography will have to rely in future on the EUV wavelength range in order to obtain a further increase in resolution and since, owing to their coating, the mirrors used in that case are capable of reflecting only about 70% of the incident light and consequently absorb about 30% of the incident light, materials having a low coefficient of thermal expansion are normally used as substrate material for such mirrors. Such so-called “low expansion materials” are, for example, Zerodur®, ULE® or Clearceram®. Those materials normally have an amorphous silicate glass content of above about 50% and, in extreme cases, of even 100%. For a projection exposure system to be capable of functioning on a long-term basis it is necessary to ensure, therefore, that the energy absorbed in the substrate material during operation does not lead to changes in the substrate and thus to degradation of the mirror surface. In other words, it is necessary to ensure that changes of any kind in the shape or roughness of the surface, which can lead to a no longer tolerable increase in aberrations or stray light, do not occur. An object of the invention to provide production methods for mirror elements so that the mirror elements exhibit no change or only a negligibly small change in surface shape on long-term irradiation with EUV. It is a further object of the invention to provide mirror elements of that kind and an EUV projection exposure system for microlithography having such mirror elements. In accordance with one aspect, that object is attained by a method for the production of a mirror element that has a reflective coating for the EUV wavelength range and a substrate, the method comprising: pre-compacting the substrate by hot isostatic pressing, and applying the reflective coating to the pre-compacted substrate, wherein either the pre-compacting of the substrate is performed until a saturation value of the compaction of the substrate by long-term EUV irradiation is reached, or, for further compaction, the pre-compacted substrate is irradiated, especially homogeneously, with ions and/or with electrons in a surface region in which the coating has been or will be applied. In accordance with the invention it is proposed that pre-compacting of a mirror substrate of an EUV mirror be carried out by hot isostatic pressing. The penetration depth of the EUV radiation impinging on the mirror element when in use, and hence the volume of the substrate in which a change in density may occur as a result of EUV irradiation, is admittedly relatively small as a rule (typically some 100 nm). The inventor has found that it is nevertheless advantageous for the entire substrate to be pre-compacted by hot isostatic pressing since such a pressure treatment can be carried out relatively quickly and inexpensively in the case of the materials used as the substrate. In one variant, there is selected as the substrate material a doped glass material or a glass ceramic, which may be selected in particular from the group comprising: ULE®, Zerodur® and Clearceram®. Such substrate materials have a low coefficient of thermal expansion which may, for example, be at most |0.5×10−7| 1/K in a range of 0° C. to 50° C. To produce such a low coefficient of thermal expansion (CTE), doped glass or glass ceramic materials are typically used—for example, as mentioned above, ULE®, Clearceram® or Zerodur®. Glass ceramic materials having the low coefficient of thermal expansion indicated above consist as a rule of a crystalline phase and a glass phase. The crystalline phase has a negative coefficient of expansion which can precisely be compensated for by the positive coefficient of expansion of the glass phase. Glass materials having a low CTE are as a rule doped glasses, for example TiO2-doped quartz glass (ULE). It will be appreciated that undoped glass, for example undoped quartz glass (fused silica), may alternatively also serve as substrate material. The materials mentioned above have an amorphous silicate glass content of more than about 50 wt. % and are therefore especially suitable for pre-compaction by hot isostatic pressing. In the hot isostatic pressing, (initial and holding) temperatures of between about 1100° C. and about 1300° C., preferably between 1150° C. and 1250° C., have proved to be especially favourable. It will be appreciated that it is not imperative for a single temperature to be maintained during the hot isostatic pressing; rather, where appropriate, it is possible, for example, for cooling to take place from a maximum temperature in several temperature stages. In a further variant, the pressure in the hot isostatic pressing is selected to be between 20 MPa and 250 MPa, preferably between 50 MPa and 150 MPa. The use of that pressure range has been found to be especially advantageous for creating a high degree of pre-compaction. In one variant, the holding time in the hot isostatic pressing is selected to be between 0.5 hour and 5 hours, preferably between 2 hours and 4 hours. It has been found that a sufficient pre-compaction can be achieved even when relatively short holding times are used. Especially in the case of materials having a high silicate glass content of more than 90 wt. % (such as in the case of ULE®), the hot isostatic pressing may be carried out substantially as described in the above-mentioned U.S. Pat. No. 6,205,818 B1, which regarding that aspect is incorporated into the present application by reference. The substrate may be compacted by the hot isostatic pressing by at least 1%, preferably by at least 1.5%, and in particular by at least 3%. Especially when the compaction is in the range of about 2%-3% or more, the saturation value of the compaction by long-term EUV irradiation can be reached and therefore the substrate, once pre-compacted, cannot be further compacted by the EUV radiation. It is, however, possible that the saturation value cannot be reached in the hot isostatic pressing or can be reached only with excessively long holding times which may, for example, be in the range of several days. For that reason, for further compaction the pre-compacted substrate may be irradiated, especially homogeneously, with ions and/or with electrons in a surface region in which the coating has been or will be applied. With this irradiation, a surface region extending from the surface of the substrate over a small depth, typically in the range of several micrometres, can be additionally compacted, so that the saturation value of the density change is reached at least in that region. In this case, the pre-compaction using the inexpensive hot isostatic pressing process permits the irradiation times of the ion or electron irradiation to be considerably reduced and thereby allows any form change possibly occurring as a result of the ion or electron irradiation to be kept as small as possible. The irradiation is advantageously carried out with high-energy ions having an energy of between 0.2 MeV and 10 MeV at a total particle density of from 1014 to 1016 ions per cm2 and/or with high-energy electrons having a dose of between 10 J/mm2 and 2000 J/mm2 at energies of between 10 KeV and 20 KeV. The irradiation may be carried out in this case especially as described in the Applicant's US 61/234815, which is incorporated in this Application by reference. It will be appreciated that, before and/or after the irradiation, additional processing steps, especially smoothing steps at the surface of the substrate, may be carried out, for example as described in the Applicant's US 61/234815. In one variant, the irradiation is carried out until there is obtained in the surface region a density that is at least 0.5%, preferably at least 1%, in particular at least 1.5% higher than the density of the remainder of the substrate. Together with the change in density obtained on pre-compaction of the substrate it is possible in this case for the saturation value of the compaction to be attained in an especially simple manner. The additionally compacted surface region extends as a rule in this case to a depth of about 5 μm from the surface of the substrate, the exact value depending on the ion or electron energy which is typically so selected that the compacted surface region extends as least as far as the penetration depth of the EUV radiation on use of the mirror. A further aspect of the invention relates to a mirror element comprising: a reflective coating for the EUV wavelength range, and a substrate, wherein the substrate is pre-compacted by hot isostatic pressing. Either the entire substrate is pre-compacted to a saturation value of the compaction of the substrate by long-term EUV irradiation, or a surface region of the substrate that extends beneath the coating has a density that is at least 0.5%, preferably at least 1%, in particular at least 1.5%, higher than that of the remainder of the substrate. The density of the pre-compacted substrate material markedly exceeds the density of the substrate material attained in a conventional production process (without pre-compaction). As explained above, the material of the substrate is typically a doped glass material or a glass ceramic having a low coefficient of thermal expansion, especially ULE®, Zerodur® or Clearceram®. Apart from the low thermal expansion of those substrate materials they have the additional advantage that they have a high silicate glass content (about 50 wt. % or more). With that material, a considerable degree of pre-compaction can be achieved by hot isostatic pressing with relatively short holding times. In one embodiment, the material of the substrate is a quartz glass doped with TiO2, in particular ULE®, an initial density of the substrate before compaction being 2.21 g/cm3. That initial density is typically obtained in ULE® that has been produced by a conventional production process. Especially when the saturation value of the density change is not yet reached by hot isostatic pressing, the resistance of the substrate to EUV irradiation can be increased if a surface region of the substrate extends beneath the coating, which surface region has a density that is at least 0.5% higher, preferably at least 1% higher, in particular at least 1.5% higher than that of the remainder of the substrate and which surface region has been obtained through high-energy ion or electron irradiation. As described above, it is advantageous if that surface region extends to a depth of about 5 μm from the surface of the substrate. By homogeneous irradiation with ions and/or electrons it is possible to achieve a homogeneous compaction of the substrate in that surface region. A further aspect of the invention is implemented in an EUV projection exposure system for microlithography, comprising an illumination system and a projection system having at least one mirror element for the EUV wavelength range as described above. In such an EUV projection exposure system, the surface shape of the mirror elements designed as described above changes on EUV irradiation only negligibly during the useful life of the system, and therefore no appreciable surface deformations that might lead to an increase in aberrations or stray light occur any longer. Further features and advantages of the invention will be apparent from the following description of illustrative embodiments of the invention with reference to the Figures of the drawings, which show details essential to the invention, and from the claims. The individual features may be implemented individually or a plurality thereof may be implemented in any desired combination in a variant of the invention. FIG. 1 shows schematically an EUV projection exposure system 1 comprising a beamshaping system 2, an illumination system 3 and a projection system 4 which are accommodated in separate vacuum housings and are arranged in succession in a beam path 6 from an EUV light source 5 of the beam-shaping system 2. A plasma source or a synchrotron, for example, may be used as the EUV light source 5. The emitted radiation in the wavelength range of between about 5 nm and about 20 nm is first collimated in a collimator 7. Using a downstream monochromator 8, the desired operating wavelength is filtered out by varying the angle of incidence, as indicated by the double-headed arrow. In the wavelength range mentioned, the collimator 7 and the monochromator 8 are usually in the form of reflective optical elements, with at least the monochromator 8 not having a multilayer system on its optical surface in order to reflect a wavelength range as broad as possible. The radiation treated in the beam-shaping system 2 in respect of wavelength and spatial distribution is introduced into the illumination system 3 which has a first and second mirror element 9, 10. The two mirror elements 9, 10 pass the radiation onto a photomask 11 forming a further mirror element that has a structure which is projected on a reduced scale onto a wafer 12 by the projection system 4. For that purpose, a third and fourth mirror element 13, 14 are provided in the projection system 4. The mirror elements 9 to 14 are in this case disposed in the beam path 6 of the EUV projection exposure system 1 and therefore are exposed to long-term EUV irradiation. As illustrated in FIG. 3 for the example of the second mirror element 10 of the illumination system 2, that mirror element consists of a reflective coating 10a applied to a substrate 10b. The reflective coating 10a is a multilayer system having alternating layers of molybdenum and silicon whose thicknesses are coordinated in such a manner that as high a reflectivity as possible is obtained at the operating wavelength of the EUV projection exposure system 1 of about 13.5 nm. To avoid deformation on heating of the mirror element 10, the substrate 10b consists of a material having a low coefficient of thermal expansion, typically doped glass or a glass ceramic, for example ULE®, Zerodur® or Clearceram®. Those materials have a silicate content of more than about 50 wt. %, the density of which becomes greater on EUV irradiation of the substrate 10b, in which case the surface of the mirror element 10 may deform, which may lead to aberrations and increased formation of stray light. To avoid compacting of the substrate 10b on long-term irradiation with the light of the EUV light source 5, a pre-compaction of the substrate 10b is carried out by placing it in a hot isostatic press 20 which is shown in FIG. 2. The press 20 is operated in this case in an inert gas atmosphere 20a, for example with argon being used as the inert gas. In the case under consideration, a sample of the substrate 10a was kept at a temperature T of about 1200° C. and at a pressure p of about 100 MPa for a period of about 4 hours, with both the heating-up and the cooling-down being performed at a rate of about 10 K/min. If, as in the example under consideration, ULE®, i.e. TiO2-doped quartz glass, having a titanium dioxide content of about 8 wt. % is used as the material for the substrate 10b, it can be compacted in the hot isostatic pressing operation using the above-mentioned parameters by about 1.5%-2% relative to its initial density of about 2.21 g/cm3. To verify the degree of compaction obtained by the hot isostatic pressing, the ULE® sample was subjected to electron irradiation at an energy dose of about 3.7×1011 rad to simulate long-term EUV irradiation. The form change obtained in that case was about 5 nm, and the degree of compaction, determined by interferometer measurement, was about 0.45%. By comparison, in the case of a non-pre-compacted ULE® sample exposed to electron irradiation at a comparable dose (4.1×1011 rad), a degree of compaction of about 1.82% was measured, with the form change being about 20 nm. Using the hot isostatic pressing, therefore, it was possible to achieve a reduction in compaction to a quarter of the value obtained in a sample that had not been pre-treated. In the case of ULE® or comparable materials, the maximum compaction (saturation value) under electron irradiation occurred as a rule above a dose of about 4 to 6×1011 rad. In silicon dioxide—and in ULE® glass also—the saturation value of the density change after long-term irradiation with EUV radiation is likewise typically on the order of about 2%-3% within the penetration depth reached by the radiation, which is typically not more than 5 μm. That saturation value may not quite be achieved by pre-compaction of the substrate 10b using the above parameter values. In order nevertheless to prevent the substrate 10b of the mirror element 10 from becoming further compacted on EUV irradiation, it is homogeneously irradiated with high-energy ions 16 in a surface region 15 in which the coating 10a has been applied. The ions 16 generally have in this case an energy of between about 0.2 MeV and about 10 MeV at a total particle density of from 1014 to 1016 ions per cm2. Alternatively, the substrate 10b may also be irradiated with high-energy electrons, the dose typically being in this case between about 10 J/mm2 and 1000 J/mm2. The irradiation with ions or electrons is carried out in this case until the density in the surface region 15 is at least 0.5%, and where applicable at least 1% or 1.5%, higher than the density of the remainder of the substrate 10b. The surface region 15 typically extends in this case at least as far as the penetration depth of the EUV radiation into the substrate 10b, which in the present case is about 5 μm. By combining the hot isostatic pressing with the irradiation, therefore, it is altogether possible for the saturation value for the density change to be attained at least in the surface region 15, so that even on long-term irradiation with the light of the EUV light source 5 the substrate 10b may experience only extremely slight surface deformation due to the change in density of the substrate 10b. It will be appreciated that the irradiation with ions and/or electrons may also be performed on the substrate 10b before coating, and that further process steps, especially smoothing of the substrate surface, may be carried out between the hot isostatic pressing and the application of the coating. For details of the irradiation of the substrate with ions or electrons, reference is again made to the Applicant's US 61/234815. It is again explicitly mentioned that, on suitable selection of the parameters of the hot isostatic pressing, where applicable also exclusively by pre-compaction of the substrate 10b, an increase in density of about 1%-3% can be achieved which, depending on the substrate material used, corresponds to the saturation value of the density change, and therefore the subsequent irradiation with ions or electrons may, where applicable, be dispensed with. In each case, mirror elements whose density remains unchanged under long-term irradiation with EUV radiation are obtained in the manner described above. The above description of the preferred embodiments has been given by way of example. From the disclosure given, those skilled in the art will not only understand the present invention and its attendant advantages, but will also find apparent various changes and modifications to the structures and methods disclosed. The applicant seeks, therefore, to cover all such changes and modifications as fall within the spirit and scope of the invention, as defined by the appended claims, and equivalents thereof. |
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046559906 | claims | 1. In a light water nuclear reactor incorporating guide tubes for the fuel assemblies of said nuclear reactors, said assemblies being immersed in operation in the cooling water of the core of such a reactor, the guide tubes being of the type made from zircaloy and fixed at their two ends respectively to an upper end part and a lower end part made from stainless steel or Inconel and which incorporate devices for braking the fall of the control rods which they house during the rapid shutdown of the reactor, the improvement wherein said braking devices comprise means for restricting the diameter of the guide tubes comprising for each guide tube a zircaloy inner sleeve spot welded to the said guide tube and whose internal diameter permits the passage, with a calibrated clearance, of the corresponding control rod, the sleeve being distributed over an area of the lower portion of each guide tube, representing approximately 1/7 of its height and associated with orifices in the actual guide tubes to produce the progressive hydraulic absorption of the end of the fall of the control rods and wherein each guide tube of the fuel assembly is fixed to the upper end part by means of a conical end bore in which it is axially locked by means of a stainless steel plug fixed in the end part. 2. In a light water nuclear reactor incorporating guide tubes for the fuel assemblies of said nuclear reactors, said assemblies being immersed in operation in the cooling water of the core of such a reactor, the guide tubes being of the type made from zircaloy and fixed at their two ends respectively to an upper end part and a lower end part made from stainless steel or Inconel and which incorporate devices for braking the fall of the control rods which they house during the rapid shutdown of the reactor, the improvement wherein said braking devices comprise means for restricting the diameter of the guide tubes comprising for each guide tube a zircaloy inner sleeve spot welded to the said guide tube and whose internal diameter permits the passage, with a calibrated clearance, of the corresponding control rod, the sleeve being distributed over an area of the lower portion of each guide tube, representing approximately 1/7 of its height and associated with orifices in the actual guide tubes to produce the progressive hydraulic absorption of the end of the fall of the control rods and wherein each guide tube of the fuel assembly is fixed to the lower end part by means of a metal plug crimped or welded to the inner sleeve and joined to the lower end part by disassemblable mechanical means. |
summary | ||
claims | 1. A method for optimizing a treatment plan for a patient for a gamma radiation therapy system, the system comprising a radiation therapy unit having a fixed radiation focus point, wherein a collimator of said therapy system is provided with a plurality of collimator passage inlets directing radiation emanating from radioactive sources of a source carrier arrangement of the therapy system to said focus point, said collimator having a plurality of sectors and wherein each sector has a number of states of collimator passage diameters which can be individually adjusted to change a spatial dose distribution surrounding the focus point, said method comprising the steps of:determining a set of shots to be delivered to a plurality of isocenter positions within a target volume of a patient during a treatment session;determining a beam-on time for each respective sector and state for each isocenter during which radiation is to be delivered based on the treatment plan; andfor each isocenter position, grouping sectors and states of respective sector in accordance with predetermined rules with respect to beam-on times for respective states of the sectors,wherein sectors and respective states are aggregated for simultaneous delivery of radiation during a predetermined period of time, and wherein the predetermined rules comprise selecting the longest beam-on time for each state of a respective sector for an aggregated simultaneous delivery of radiation and wherein the predetermined period of time for simultaneous delivery of radiation is determined to be the minimum beam-on time for a state of a sector of the aggregated sectors. 2. The method according to claim 1, further comprising the steps of:checking whether the minimum beam-on time for a state of a sector of the aggregated sectors is zero;setting the sector having a zero beam-on time to be blocked, wherein the sector is blocked for delivery of radiation; andselecting the shortest non-zero beam-on time as the period of time for delivery of radiation. 3. The method according to claim 2, further comprising the step of excluding a set of sectors and respective states aggregated for a simultaneous delivery of radiation during a predetermined period of time from delivery if said predetermined period of time is below a predetermined threshold. 4. The method according to claim 1, further comprising the step of excluding a set of sectors and respective states aggregated for a simultaneous delivery of radiation during a predetermined period of time from delivery if said predetermined period of time is below a predetermined threshold. 5. A sector planning module for a gamma radiation therapy system, the system comprising a radiation therapy unit having a fixed radiation focus point, wherein a collimator of said therapy system is provided with a plurality of collimator passage inlets directing radiation emanating from radioactive sources of a source carrier arrangement of the therapy system to said focus point, said collimator having a plurality of sectors and wherein each sector has a number of states of collimator passage diameters which can be individually adjusted to change a spatial dose distribution surrounding the focus point, said sector planning module comprising:a dose distribution module adapted to obtain information of a set of shots to be delivered to a plurality of isocenter positions within a target volume of a patient during a treatment session, and of a beam-on time for each respective sector and state for each isocenter during which radiation is to be delivered based on the treatment plan; andan aggregation module adapted to, for each isocenter position, group sectors and states of respective sector in accordance with predetermined rules with respect to beam-on times for respective states of the sectors,wherein sectors and respective states are aggregated for simultaneous delivery of radiation during a predetermined period of time and wherein the predetermined rules comprise to select the longest beam-on time for each state of a respective sector for an aggregated simultaneous delivery of radiation and wherein the predetermined period of time for simultaneous delivery of radiation is determined to be the minimum beam-on time for a state of a sector of the aggregated sectors. 6. The sector planning module according to claim 5, whether said aggregation module is adapted to:check whether the minimum beam-on time for a state of a sector of the aggregated sectors is zero;set the sector having a zero beam-on time to be blocked, wherein the sector is blocked for delivery of radiation; andselect the shortest non-zero beam-on time as the period of time for delivery of radiation. 7. The sector planning module according to claim 6, wherein said aggregation module is adapted to exclude a set of sectors and respective states aggregated for a simultaneous delivery of radiation during a predetermined period of time from delivery if said predetermined period of time is below a predetermined threshold. 8. The sector planning module according to claim 5, wherein said aggregation module is adapted to exclude a set of sectors and respective states aggregated for a simultaneous delivery of radiation during a predetermined period of time from delivery if said predetermined period of time is below a predetermined threshold. |
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description | This application is the national stage of International Application No. PCT/KR2013/012064, filed Dec. 24, 2013, entitled, “Simulation Construction Method For The Measurement Of Control Rod Insertion Time,” which claims the benefit of priority of Korean Patent Application No. 10-2012-0155251, filed Dec. 27, 2012, the contents of both of which are incorporated herein by reference in their entirety. One or more embodiments of the present invention relate to a method of constructing a simulation to measure a control rod insertion time, and more particularly, to an evaluation methodology for the measurement of the insertion time of a control rod which freely falls down to the nuclear reactor core under gravity by use of the computational fluid dynamics (CFD). One or more embodiments of the present invention relate to a method of calculating a value similar to a real value of free-falling of a control rod in a nuclear power-plant considering the three-dimensional thermal hydraulic effect in a nuclear reactor. A control rod is a core component with a function of controlling core reactivity by changing the number of neutrons in a core while inserting a control rod into the core and retracting a control rod out of the core. In other words, as a component used with a purpose of controlling and halting the power of a nuclear reactor, the control rod is moved up and down by a control rod drive mechanism installed on the top portion of the nuclear reactor. After receiving a fall signal from the control rod drive mechanism, the control rod freely falls down to the nuclear reactor core with the gravity. The insertion time of a control rod absolutely needs evaluation for the safe operation of a nuclear reactor and satisfaction of the allowance requirements. A conventional control rod insertion time is calculated using one-dimensional codes. The one-dimensional codes consider factors such as hydraulic resistance of a fluid, friction, and weight of a control rod assembly, but have a drawback of a big calculation error due to no consideration of three-dimensional hydraulic effects occurring inside a guide tube for guiding the control rod. Therefore, the conventional control rod insertion time has been conservatively evaluated by applying the high uncertainly of a value bigger than a real design. Under these circumstances the control rod insertion time may be much larger than measured time during the actual operation of the nuclear reactor. Since the control rod insertion time is used to perform the operation of the nuclear reactor and the safety analysis evaluation, a practical value needs to be deduced with elimination of exaggerated conservatism. To evaluate the control rod insertion time similar to the actually measured value in the nuclear reactor, the thermal-hydraulic phenomenon having a three-dimensional, spatial distribution of factors such as the control rod, the nuclear reactor, the guide tube, and an impact absorption tube needs to be taken into consideration. One or more embodiments of the present invention include a method of constructing a simulation to measure a control rod insertion time, by preparing a three-dimensional model for the free-falling time of the control rod with a real configuration for surrounding structures such as a nuclear reactor, a control rod, a guide tube, an impact absorption tube, and a drainage, and by simulating a three-dimensional thermal-hydraulic phenomenon using the computational fluid dynamics (CFD), thereby producing a value of the control rod insertion time that is similar to the time actually measured in the nuclear reactor. According to one or more embodiments of the present invention, the simulation construction method for the measurement of control rod insertion time may include the three-dimensional modeling, when the control rod freely falls down to the core of the nuclear reactor by the gravity, of an inside wall of the nuclear reactor, the control rod accommodated into the nuclear reactor, the guide tube guiding the control rod, the impact absorption tube installed at the bottom of the guide tube with a diameter smaller than the diameter of the guide tube, the first flow hole located on the wall side of the guide tube, and the second flow hole located at the bottom of the impact absorption tube; the construction of the flow field by dividing the inside of the nuclear reactor into multiple number of cells; the construction of the flow field formed by a variable grid system with multiple number of changing cells where the configuration changes as the location of the control rod changes on the moving path of the control rod and by an aligned grid system with multiple number of fixed cells where the configuration is maintained regardless of the location change of the control rod; the calculation of the simulation estimated value for the insertion time until the control rod is inserted into the impact absorption tube by calculating the thermal-hydraulic phenomenon using the three-dimensional CFD codes; and the cell adjustment where the estimated value is compared with the actual value of the insertion time when the control rod actually falls inside the nuclear reactor, and the error between the estimated value and the actual value is verified if within an arbitrarily reference range, and, when the error exceeds the reference range, the size of the variable cell and/or the size of the fixed cell is changed. The simulation construction method for the measurement of control rod insertion time may include the calculation of the hydraulic drag of the control rod caused by the water filling the nuclear reactor, when the control rod falls. The simulation construction method for the measurement of control rod insertion time may include the consideration of the friction of the water and the control rod, and the weight of the control rod, when the control rod falls. The simulation construction method for the measurement of control rod insertion time may include the establishment of the reference range in such a way that the estimated value and the actual value make an error of less than 5%. The simulation construction method for the measurement of control rod insertion time may include the consideration of the thermal-hydraulic phenomenon in such a way that the pressure change, the temperature change, and the density change of the water around the control rod. As described above, according to the one or more of the above embodiments of the present invention, the simulation construction method for the measurement of the insertion time of the control rod may exhibit an effect that utilization of the CFD and the realization of the falling process of the control rod by considering the three-dimensional thermal-hydraulic characteristics may obtain the estimated insertion time of the control rod calculated by the simulation as a very approximate value for the actual value when the control rod actually falls inside the nuclear reactor. By simulating the control rod insertion time as an approximate value for the actual value, the additional information for the drop speed of the control rod, the pressure distributions in the guide tube and the impact absorption tube, and the flow rates out of the first and second flow holes may be obtained to be utilized as useful information for design and operation of the nuclear reactor. As described above, according to the one or more of the above embodiments of the present invention, the simulation construction method for the measurement of the insertion time of the control rod may exhibit an effect that the information obtained out of the present invention may be applied to design and operation of not only new but currently-operating nuclear reactors, and provide useful information for the operation of the nuclear reactor currently in operation. Reference will now be made in detail to embodiments, examples of which are illustrated in the accompanying drawings, wherein like reference numerals refer to the like elements throughout. In this regard, the present embodiments may have different forms and should not be construed as being limited to the descriptions set forth herein. Accordingly, the embodiments are merely described below, by referring to the values, to explain aspects of the present description. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. Hereinafter, the present invention will be described in detail by explaining exemplary embodiments of the invention with reference to the attached drawings. In order to further clearly describe features of one of more embodiments of the present invention, well-known technologies are not particularly described herein. A control rod begins free-falling toward a core of a nuclear reactor by gravity after receiving a falling signal from a control rod drive mechanism. The present invention relates to a method of constructing a simulation to measure a time from when the control rod begins free-falling until the control rod enters an impact absorption tube. FIG. 1 is a flowchart of a simulation construction method according to an embodiment of the present invention. FIG. 2 illustrates a falling process of a control rod. FIGS. 3 and 4 schematically illustrate a variable grid system and an aligned grid system according to the position of a control rod. Referring to FIG. 1, according to an embodiment of the present invention, the simulation construction method for measurement of the control rod insertion time may include a three-dimensional modeling operation for the geometric shape, a flow field forming operation, a simulation estimated value calculation operation, and a cell change operation. The three-dimensional modeling operation may be the three-dimensional modeling of surrounding structures where a control rod 10 freely falls. According to the present embodiment, an inside wall 20 of a nuclear reactor, the control rod 10, a guide tube 30, an impact absorption tube 40, a first flow hole 50, and a second flow hole 60 may be modeled in three dimensions. The nuclear reactor may accommodate the control rod 10, and the guide tube 30 may guide the control rod 10 when the control rod 10 freely falls. The control rod 10 falls downward along the guide tube 30. The impact absorption tube 40 may be installed at the bottom of the guide tube 30 and have a diameter smaller than the diameter of the guide tube 30. The control rod 10, while the control rod 10 falls along the guide tube 30, may exhibit a rapid acceleration until the control rod 10 arrives at the impact absorption tube 40. At the time when the control rod 10 enters the impact absorption tube 40 the control rod 10 may experience an increase of the hydraulic drag and the drop speed of the control rod 10 decreases. In other words, since the diameter of the impact absorption tube 40 may be arranged as smaller than the diameter of the guide tube 30, the drop speed of the control rod 10 may decrease when the control rod 10 passes through the impact absorption tube 40. The first flow hole 50 may be installed on the wall of the guide tube 30. When the control rod 10 falls along the guide tube 30, the control rod 10 may push the water filling inside of the guide tube 30, and the water may move to the outside of the guide tube 30 through the first flow hole 50. In other words, the water may fill the entire nuclear reactor and, when the control rod 10 falls, the water existing inside the guide tube 30 may move to the outside of the guide tube 30 through the first flow hole 50. The second flow hole 60 may be installed at the bottom of the impact absorption tube 40. When the control rod 10 may fall further downward after entering the impact absorption tube 40, the water filling the inside of the impact absorption tube 40 may move to the outside through the second flow hole 60. In other words, the water may move to the outside of the impact absorption tube 40 through the second flow hole 60. The three-dimensional modeling operation may include the modeling of surrounding structures related with the control rod 10 in three dimensions while the control rod 10 falls. A flow field may be formed. The flow field may include a variable grid system 80 and an aligned grid system 70. The flow field may signify an area influenced by the falling of the control rod 10 while the control rod 10 falls inside of the nuclear reactor, and is constructed by dividing the inside of the nuclear reactor into multiple cells. For the measurement of the insertion time of the control rod 10, a core area may be divided into many small cells. The small cells are defined as grids. The variable grid system 80 may be a set of multiple cells having a shape that changes as the location of the control rod 10 changes on a path along which the control rod 10 passes. Each cell with a shape that changes as the location of the control rod 10 changes may be referred to as a variable cell 81. In other words, the variable grid system 80 is a set of the variable cells 81 which are arranged on the path along which the control rod 10 passes while the control rod 10 falls. The variable grid system 80 may be generally obtained by performing calculations with changing the grid at every predetermined time step when a subject exists moving relatively to the flow field or the configuration of the flow field changes due to the influence of certain variables in the flow field. In other words, according to the present embodiment, since the grid of a moving area where the control rod 10 falls as the control rod 10 falls toward the core due to the gravity, may change along with the control rod 10 as the time passes, the variable cell 81 may be configured as changing according to the location of the control rod 10. According to the present embodiment, the variable grid system 80 may be arranged as illustrated in FIGS. 3 and 4. FIG. 3 illustrates the top portion of the control rod 10, in which the variable grid system 80 that is affected according to a change in the position of the control rod 10 is configured from the point where the top portion of the control rod 10 is located before the fall to the point where the top portion of the control rod 10 is located after the fall. The variable grid system 80 is constructed in this range by use of the variable cell 81. FIG. 4 illustrates the bottom portion of the control rod 10, in which the path where the bottom portion of the control rod 10 passes as the control rod 10 falls is configured as the variable grid system 80. The aligned grid system 70 may be configured by dividing the area in the flow field other than the variable grid system 80 into multiple cells. The aligned grid system may be aggregation set of multiple fixed cells 71 maintaining the configuration regardless of the location change of the control rod 10. Each fixed cell 71 included in the aligned grid system 70 may have a rectangular shape to be regularly distributed. A non-aligned grid system, a term in contrast with the aligned grid system 70, conventionally refers to an irregular arrangement of non-aligned grids which have cells with a shape other than a rectangle; for example, a quadrangle without sides having a right angle, or a triangle. In the present embodiment, the aligned grid system 70 may be used for the calculation accuracy and the reduction in the calculation time, and the area where the control rod 10 moves may be configured by the variable grid system 80. The simulation estimated value calculation operation may be an operation of calculating an insertion time until the control rod 10 is inserted into the impact absorption tube 40 by analyzing the thermal-hydraulic phenomenon by use of the three-dimensional computational fluid dynamics (CFD) codes. The thermal-hydraulic phenomenon may indicate physical changes occurring around a subject when the subject falls and, according to an embodiment of the present invention, includes factors such as the pressure change, the temperature change, and the density change of the water around the control rod 10 while the control rod 10 falls down to the impact absorption tube 40. The thermal-hydraulic phenomenon using the three-dimensional CFD codes may be analyzed in consideration of variable factors such as pressure, temperature and density of the water filling the nuclear reactor. Since the analysis method of the thermal-hydraulic phenomenon by use of the three-dimensional CFD codes is a well-known method, the analysis method will not be particularly described herein. The cell configuration change operation may be performed to reduce an error range by changing the size of the variable grid cell 81 and/or the fixed cell 71 when an error between the estimated value and the actual value taken until the control rod 10 enters the impact absorption tube 40 in a real power plant after the control rod 10 begins to fall exceeds the error range. The estimated value may be compared with the actual value for the insertion time when the control rod 10 falls in a real power plant. Then, it is determined whether an error between the estimated value and the actual value is within a certain reference range. According to an embodiment of the present invention, the reference range is set in such a way that the error between the estimated value and the actual value is less than 5%. When the reference range is set to exceed 5%, the evaluation result of the estimated value may be over-conservative, which is not desirable. When the error exceeds the reference range, the change operation of the cell configuration may be performed to change the size of the variable cell 81 and/or the fixed cell 71. As used herein, the term “and/or” includes either a simultaneous size change of both the variable cell 81 and the fixed cell 71, or a change of a size of any selected one of either the variable cell 81 or the fixed cell 71. An accurate estimated value may be calculated by optimizing the mesh density of the variable cell 81 included in the variable grid system 80 and/or the fixed cell 71 included in the aligned grid system 70, since the structures arranged around the control rod 10 are very complicated and elaborate when the control rod 10 falls. According to an embodiment of the present invention, the simulation construction method for measurement of the control rod insertion time may additionally include the calculation operation of the hydraulic drag, the friction between the water and the control rod 10, and the consideration operation of the weight of the control rod 10. The calculation operation of the hydraulic drag may consider the influence of the control rod 10 on the hydraulic drag, while the control rod 10 falls, due to the water filling inside the nuclear reactor, and may more accurately produce the estimated value by considering the actual drag by the water as a parameter in the estimation calculation, along with the simulation of the three-dimensional thermal-hydraulic phenomenon. When the control rod 10 falls down by the gravity to the impact absorption tube 40, the water vertically rises along a narrow flow path formed by the guide tube 30 and the impact absorption tube 40. In such a case when the water vertically moves, the hydraulic drag may be obtained by using Darcy's equation. Since the detailed process of calculating the hydraulic drag by applying Darcy formula is well known, and the process of calculation is outside the range of the technical concept of the present invention, the detailed process thereof will not be particularly described herein. The operation of considering the friction and the weight of the control rod 10 may work as another parameter to calculate the estimated value for the insertion time of the control rod 10. The friction may signify a frictional force generated by a viscosity coefficient of water and a surface roughness of the inside wall of the guide tube 30 when the control rod 10 falls along the guide tube 30 filled with the water and the impact absorption tube 40. In other words, when the control rod 10 falls, the friction is generated in a direction opposite to the direction in which the control rod 10 falls, on the surface of the control rod 10 in contact with the water. Since the friction acts in the direction opposite to the falling direction of the control rod 10, the friction acts as a factor delaying the insertion time of the control rod 10. Therefore, the friction is considered to calculate an accurate estimated value. The weight of the control rod 10 may be used to calculate the estimated value by applying the weight of the control rod 10 actually used in a real power plant as a parameter. As described above, the simulation construction method for measurement of the control rod insertion time according to the present invention may exhibit an effect that the estimated value for the insertion time of the control rod 10 is predicted as close to the actual insertion time of the control rod 10 which actually falls in the nuclear reactor by simulating the process of the control rod 10 which freely falls from the top portion of the nuclear reactor down to the impact absorption tube 40. As described above, the simulation construction method for measurement of the control rod insertion time according to the present invention may exhibit an effect that the accuracy of the estimated value is more enhanced by considering the hydraulic drag, the friction, and the weight of the control rod 10 as parameters for calculating the estimated value. As described above, the simulation construction method for measurement of the control rod insertion time according to the present invention may exhibit an effect that useful information for the design and the operation of the nuclear reactor is provided by additionally obtaining the information on the drop speed of the control rod 10, the pressure distribution inside the guide tube 30 and the impact absorption tube 40, and a flow rate at the first and second flow holes 50 and 60 when the control rod 10 falls, based on the estimated value close to the actual value. It should be understood that the exemplary embodiments described therein should be considered in a descriptive sense only and not for purposes of limitation. Descriptions of features or aspects within each embodiment should typically be considered as available for other similar features or aspects in other embodiments. While one or more embodiments of the present invention have been described with reference to the values, it will be understood by those of ordinary skill in the art that various changes in form and details may be made therein without departing from the spirit and scope of the present invention as defined by the following claims. |
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058754075 | summary | BACKGROUND OF INVENTION 1. Field of Invention The present invention relates to a method for immobilizing waste chlorides salts containing radionuclides and hazardous nuclear material for permanent disposal, and, in particular, a method for immobilizing waste chloride salts containing cesium, in a synthetic form of pollucite. 2. Description of Related Art Electrorefining methods involving electrochemical cells are used for the recovery of fissionable materials from spent nuclear reactor fuels, including uranium and plutonium. Typically, in an electrorefining cell, an electrolyte consisting of a molten eutectic salt mixture, such as KCl and LiCl, is used to transport the metal or metals to be purified between electrode solutions. When used to treat spent nuclear reactor fuels, the salt mixture becomes contaminated with radionuclides (e.g., .sup.-137 cesium and .sup.-90 strontium), hazardous materials (e.g., barium), and other species (e.g., sodium and .sup.-129 iodine). Eventually, the salt mixtures are no longer suitable for use in the electrorefining cell. Since the separation of cesium and strontium from the salt is difficult, the cesium and strontium, and any other radionuclides and toxic metal chlorides and iodides, are disposed along with a portion of the salt matrix. The waste salt containing the cesium and strontium is a high level waste (HLW) which must be deposited in an HLW geologic repository. To prevent an uncontrolled release of the radionuclides and other hazardous chemicals into the groundwater, the waste form must be leach resistant. Due to the very high water solubility of the waste salts, a method for encapsulating and immobilizing the waste salt is required. The high solubility and volatility of cesium, in particular, has caused difficulties in the identification and preparation of suitable waste forms which would immobilize cesium for the necessary, extended storage time. For example, incorporation of cesium into traditional solid waste forms, including borosilicate glass, synroc, cement, or ceramics, is ineffective due to the relatively high leach rates resulting from the inherently high solubility of cesium and difficulties during processing requiring additional steps. For example, immobilizing cesium in a glass matrix involves converting the waste chloride salts into oxides or other chemical forms compatible with the glass-making process. These conversion processes are expensive and time-consuming, and restructuring the materials to form the glass requires temperatures of about 1000.degree. C., which increases the risk of volatilizing the cesium. Ion-exchangers are an important class of materials used for the immobilization of cesium in salt solutions and molten salts. Certain cation exchange resins and various cation exchangers, such as naturally occurring and synthetic zeolites (tectosilicate mineral), are available for selectively recovering cesium from contaminated solutions. For example, zeolite matrices, including zeolite A, zeolite X, and chabazite, have been used to immobilize waste chloride salts containing cesium because of their sorption and ion exchange properties. U.S. Statutory Invention Registration H1227 discloses contacting molten waste chloride salt containing cesium with dehydrated zeolite A and maintaining the contact to allow the salt to penetrate the zeolite cavities. The salt is occluded in the zeolite and the cesium in the non-occluded salt is sorbed by ion-exchanging with the cations in the zeolite or the occluded salt, resulting in a leach resistant aluminosilicate matrix, wherein the cesium ions are present as either aluminosilicates or as occluded salt molecules. U.S. Pat. No. 4,808,318 further describes the use of a hydrated sodium phlogopite mica (phyllosilicate mineral) to recover cesium ions from waste solutions, whereby the cesium is selectively absorbed by the modified phlogopite and fixed for long-term storage. Although methods for immobilizing cesium using ion-exchange materials can effectively purify the salt, the non-occluded surface salt must be removed from the ion-exchanger before it can be stored. In addition, if the cesium is in the form of dry, solid cesium chloride, the cesium chloride must first be dissolved in solution before being ion exchanged into a zeolite. Impurities in the solution, such as the presence of competing sodium and/or potassium ions, may also decrease the zeolite capacity for cesium. Problems are encountered in making dense, leach-resistant waste forms directly from the salt-occluded waste product, and further steps are generally required to immobilize the cesium in the ion-exchange matrix, including calcination at high temperatures or incorporation of either the cesium ions eluted from the zeolite or the zeolite containing the cesium ions into a storage medium, such as glass or cement. A preferred method of disposing of radionuclides is by encapsulation in specific crystalline, mineral waste forms, such as sodalite and pollucite, because of their refractory properties and high resistance against leaching. Generally, these methods include mixing the radioactive ions with inorganic materials and applying heat and/or pressure to form the synthetic mineral. U.S. Pat No. 5,340,506 discloses forming a sodalite intermediate from alumina, silica, and sodium hydroxide, and mixing the sodalite intermediate with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides. The mixture is compacted under heat and pressure conditions to form sodalite, whereby the waste salt and radionuclides are trapped within the sodalite cage structure. U.S. Pat. No. 5,613,240 further discloses a method for producing sodalite from salt occluded zeolites by the use of heat or heat and pressure in the presence of glass. The method involves heating substantially dry zeolite, waste chloride salts, and glass to a temperature up to about 725.degree. C. to convert the zeolite to sodalite, and thereafter maintaining the sodalite at a pressure and temperature sufficient to form a sodalite product. Pollucite (CsAlSi.sub.2 O.sub.6), a naturally occurring aluminosilicate containing cesium, is known to be one of the best crystalline waste forms for the containment of radioactive cesium, and, therefore, is a component of several mineral-based nuclear waste forms, such as synroc. Pollucite has a high loading of cesium, high thermal stability, and a high resistance to leaching. Synthetic pollucite has been produced for immobilizing radioactive ions by aqueous hydrothermal processes and high temperature synthesis, both involving an ion exchange step. Hydrothermal transformation and recrystallization of zeolites, specifically zeolite A, zeolite X, and zeolite Y, which have been loaded with cesium in an ion exchange step, occurs in the presence of water vapor at temperatures and pressures between about 260.degree. C. to 300.degree. C. and 10 to 30 MPa, respectively. However, cesium loaded chabazite subjected to hydrothermal processes reportedly results in the formation of zeolite and not a new mineral phase, such as pollucite. Pollucite has also been produced by the hydrothermal reaction of cesium with siliceous sinter in a sodium hydroxide solution. High temperature synthesis involves subjecting ion-exchanged and cesium loaded zeolite to calcination at temperatures of about 1,200.degree. C. to produce synthetic pollucite. Dense waste forms are further produced by calcination followed by sintering or by hot pressing. U.S. Pat. No. 5,591,420 discloses a new material: cesium titanium silicate pollucite (CsTiSi.sub.2 O.sub.6.5 or Cs.sub.2 Ti.sub.2 Si.sub.4 O.sub.13), which represents a new class of crystalline phase of Ti-containing zeolites, wherein the cages formed within the compound trap the cesium ions. The method of making the silicotitanate pollucite involves a one-step, direct thermal conversion at low temperatures (700.degree. C. to 1000.degree. C.), which minimizes the risk of volatilizing the cesium and reduces waste volumes. The cesium titanium silicate materials are made by selecting and combining proportions of cesium, silica, and titania, and heat treating the mixture. The components can be combined either by mixing oxides or carbonates of cesium, titanium, and silicon, or by synthesizing and hydrolyzing precursor materials. The resulting compounds are durable glass and ceramic materials, exhibiting low leach rates. Problems associated with the current methods of synthesizing pollucite are high temperature requirements, which may cause volatilization and loss of the cesium, numerous process steps that increase the cost of immobilizing the cesium, and insufficient leach resistance of the storage material. In addition, cesium titanium silicate pollucite is thermally unstable and difficult to produce on a large scale. A need exists for a method for immobilizing cesium for long term storage by incorporating the cesium into synthetic pollucite which overcomes the problems experienced in the prior art. The present method is a simple, one step conversion process for synthesizing pollucite that eliminates the need for aqueous ion exchange and/or hydrothermal synthesis at elevated temperatures and pressures. This reduces the number of waste streams. Since the solid cesium chloride is converted to pollucite without the need to dissolve the cesium chloride in solution to perform an ion exchange step, this method can be used in conjunction with other dry processing technologies, such as pyroprocessing. The present method for synthesizing pollucite includes mixing dry, non-aqueous cesium chloride with chabazite and heating the mixture to a temperature greater than the melting temperature of the cesium chloride (approximately 680.degree. C.), or above about 700.degree. C. The unexpected and surprising result is that pollucite forms in the presence of the chloride ion. In particular, the chloride appears to remain in the structure of the pollucite, apparently occluded as sodium chloride. The method significantly improves the rate of retention of cesium in ceramic products comprised of a salt-loaded zeolite by adding about 10% chabazite by weight to the salt-loaded zeolite prior to conversion at elevated temperatures and pressures to the ceramic composite. Therefore, in view of the above, a basic object of the present invention is to provide a simpler method for immobilizing radioactive cesium for long term storage. A further object of this invention is to provide a method for immobilizing radioactive cesium eliminating the need for hydrothermal synthesis which involves an aqueous ion exchange step for isolation of the cesium ion. In addition, the method is practiced at low temperatures, reducing the risk of loss of cesium. Another object of this invention is to provide a method for immobilizing radioactive cesium by mixing non-aqueous cesium chloride with chabazite and heating the mixture to a temperature sufficient to form pollucite. Yet another object of the invention is to provide a method for immobilizing radioactive cesium in a glass product including mixing non-aqueous cesium chloride with chabazite, heating the mixture to a temperature sufficient to form pollucite, cooling the pollucite, and heating the pollucite with glass to a temperature sufficient to form a glass pollucite product. Yet another object of the invention is to provide a method for immobilizing radioactive cesium which includes mixing a predetermined amount of chabazite with cesium chloride and zeolite A to form a mixture of aluminosilicates, including pollucite. Yet another object of the invention is to significantly improve the rate of retention of cesium in ceramic products comprised of a salt-loaded zeolite by adding about 10% chabazite by weight to the salt-loaded zeolite prior to its conversion at elevated temperatures and pressures to the ceramic composite. 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 instrumentation and combinations particularly pointed out in the appended claims. BRIEF SUMMARY OF THE INVENTION The present invention relates to a method for immobilizing waste chloride salts containing radionuclides and hazardous nuclear material for permanent disposal, and, in particular, a method for immobilizing waste chloride salts containing cesium, in a synthetic form of pollucite. Briefly, the method for synthesizing pollucite from chabazite and cesium chloride includes mixing dry, non-aqueous cesium chloride with chabazite and heating the mixture to a temperature greater than the melting temperature of the cesium chloride, or above about 700.degree. C. The unexpected and surprising result of the method is that pollucite forms in the presence of the chloride ion. In particular, the chloride remains in the structure of the pollucite, apparently occluded as sodium chloride. Thus, the ion exchange step, wherein cesium is separated from the chloride ion, prior to conversion to the waste product is unnecessary. In an alternate embodiment, the method for synthesizing pollucite from chabazite and cesium chloride includes mixing dry, non-aqueous cesium chloride with zeolite A to form a pollucite/sodalite product. The method, further, significantly improves the rate of retention of cesium in ceramic products comprised of a salt-loaded zeolite by adding about 10% chabazite by weight to the salt-loaded zeolite prior to its conversion at elevated temperatures and pressures to the ceramic composite. |
description | This application claims the benefit under 35 USC 119(e) of U.S. provisional application Ser. No. 61/583,852 filed Jan. 6, 2012. Radiation Shielding Barriers, also known as Radiation Protection Barriers Generally, a room designed to shield against ionizing radiation e.g.: X-Rays/Gamma Rays will require a continuous lead barrier, in the walls to a height of 2100 mm or higher that is designed to attenuate the ionizing radiation being emitted from the imaging equipment installed in the room. Generally the current method used in the attempt to provide a lead radiation shielding barrier is to construct a metal stud partition, installing to one side of the gyproc, a layer of sheet lead laminated to the back face. The laminated gyproc's manufacturers' installation instructions direct the installer to screw the laminated gyproc to the studs through predrilled fastening holes, then countersink all of the fastening screw heads, minimum 3 mm, through the face paper allowing for attachment of lead screw caps over, screwing the laminated gyproc to the studs through pre-drilled fastening holes. This act of countersinking the screws into the gypsum may destroy the structural integrity of the gypsum board. Additionally, once the screws are countersunk in the gypsum board, lead screw caps or buttons are hammered into position over every screw. This may be costly in both construction time and labour cost. Many caps/buttons are not properly installed or not installed at all allowing radiation leaks. Moreover in the angular incidence there is not enough steel in the drywall screws to provide attenuation equivalent to that provided by the lead. Furthermore, in addition to lead screw caps, a strip of sheet lead equaling the attenuating capability of the lead laminated gypsum board must be installed between the leaded face of the gypsum board and the face of the metal stud where two or more sheets of gypsum board are butted together, to prevent leakage at this seam. This may be costly in both construction time and labour cost. Many joining strips are left out allowing large gaps for ionizing radiation leaks. Other leak-proof barrier systems were allegedly created without the lead screw caps being installed over every screw or the strip of sheet lead designed to provide radiation shielding not being installed at the seams. U.S. Pat. No. 6,550,203 to Little suggests that when a lead barrier plate is installed which extends beyond the flanges of the metal stud, the countersunk holes and the lead strip added at the juncture where two or more sheets of lead laminated gypsum board abut, are no longer required. U.S. Pat. No. 4,038,553 to McCullagh discloses a clamping apparatus having a stud capturing portion and a stud facing portion bolted together to clamp a lead sheet. All previously described methods of creating a radiation leak proof lead barrier systems require extensive time consuming skilled labour to install. In addition, the integrity of the lead in some previously described systems is compromised by requiring fixed sizes of components that are not adjustable to accommodate different thicknesses or weights of lead. In addition, the lead attachment method described in some previously described systems is compromised by way of the system itself being difficult to access, difficult to install or too unclear for installers during installation procedures. In addition, the lead attachment method described in some previously described systems is compromised by way of the system itself failing to meet the guidelines for the structural support of lead sheets set out by the International Lead Association (London). In addition, the installation of any electrical devices, such as for example, switches or receptacles; or plumbing fixtures, such as water supply and drainage lines, that are required to be installed in walls, would require either damaging or penetrating the lead barrier or would not allow the barrier to be installed at all. There is provided in an embodiment a radiation shielding barrier wall assembly comprising wall elements including studs and support bars attached to and extending between the studs, sheets of radiation shielding material suspended from the support bars; and adjacent sheets of radiation shielding material overlapping on either side of a portion of a stud. In various embodiments, there may be included any one or more of the following features: the radiation shielding barrier wall assembly may have at least a stud which is elongated in a first direction, and a cross section in the plane perpendicular to the first direction having bends and legs between the bends, the cross section of the stud comprising a first leg at one end of the cross section and a second leg at another end of the cross section and further comprising an offset leg located between the first leg and the second leg, the offset leg being arranged to allow a sheet of radiation shielding material supported by a wall element to one side of the stud to extend adjacent to the offset leg to overlap with a sheet of radiation shielding material supported adjacent to the offset leg by a wall element to the other side of the stud. The radiation shielding barrier wall assembly may also have a restraining bar attached to the at least a stud to hold the sheet of radiation shielding material extending adjacent to the leg. The radiation shielding barrier wall assembly may also have finishing material attached to the studs and enclosing the sheets of radiation shielding material. The studs may extend between a top track and a bottom track. The radiation shielding barrier wall assembly may also have a corner assembly for a bend in the wall, the corner assembly comprising radiation shielding material attached such as by lamination to sheet metal, the corner assembly being attached to a stud at one side of the corner assembly, and attached to another stud at an opposite side of the corner assembly, and the corner assembly having a bend between the side and the opposite side. There is provided in an embodiment a radiation shielding barrier wall assembly comprising studs, support bars attached to and extending between the studs, and sheets of radiation shielding material suspended from the support bars. In another embodiment, there is provided a radiation shielding barrier wall assembly comprising wall elements including studs, sheets of radiation shielding material supported by the wall elements and adjacent sheets of radiation shielding material overlapping on either side of a portion of a stud. In various embodiments, there may be included any one or more of the following features: the radiation shielding barrier wall assembly may have at least a stud which is elongated in a first direction, and has a cross section in the plane perpendicular to the first direction having bends and legs between the bends, the cross section of the stud comprising a first leg at one end of the cross section and a second leg at another end of the cross section and further comprising an offset leg located between the first leg and the second leg, the offset leg being arranged to allow a sheet of radiation shielding material suspended from a support bar to one side of the stud to extend adjacent to the leg to overlap with a sheet of radiation shielding material suspended from a support bar to the other side of the stud. The radiation shielding barrier wall assembly may also have a restraining bar attached to the at least a stud to hold the sheet of radiation shielding material extending adjacent to the leg. The radiation shielding barrier wall assembly may also have finishing material attached to the studs and enclosing the support bars and sheets of radiation shielding material. The studs may extend between a top track and a bottom track. The sheets of radiation shielding material may each have a foldover edge which extends over the top of a support bar when the respective sheet of radiation shielding material is suspended from the respective support bar. The radiation shielding barrier wall assembly may also have a corner assembly for a bend in the wall, the corner assembly comprising radiation shielding material attached such as by being laminated to sheet metal, the corner assembly being attached to a stud at one side of the corner assembly, and attached to another stud at an opposite side of the corner assembly, and the corner assembly having a bend between the side and the opposite side. The sheets of radiation shielding material may comprise lead panels. In an embodiment there is provided a kit for producing a radiation shielding barrier wall assembly, the kit comprising wall elements including studs and support bars attached to and extending between the studs, the support bars configured to suspend sheets of radiation shielding material, and the studs configured to accept sheets of radiation shielding material overlapping on either side of portions of a stud. In an embodiment there is provided a stud for use in a radiation shielding barrier wall assembly, the stud being elongated in a first direction, and having a cross section in the plane perpendicular to the first direction having bends and legs between the bends, the cross section of the stud comprising a leg generally in the middle of the cross section of the stud, the leg arranged to allow a sheet of radiation shielding material supported to one side of the stud to extend adjacent to the offset leg to overlap with a sheet of radiation shielding material supported adjacent to the offset leg to the other side of the stud. In a further embodiment the stud may also have at least a hole to attach a support bar to the stud for suspending a sheet of radiation shielding material. Lead sheets may be color coded according to the thickness of the lead sheets. These and other aspects of the device and method are set out in the claims, which are incorporated here by reference. The following objects apply to one or more embodiments, but may not apply to all embodiments. It is an object in at least an embodiment of the invention to provide a radiation shielding barrier without the need for lead lined drywall, lead buttons, and lead strips. Another object in at least an embodiment is to permit elimination of the need for an installer specializing in lead installations only and allow the system to be installed by other skilled trades. Another object in at least an embodiment is to reduce the construction time of shielded rooms. Another object in at least an embodiment is to allow for easy installation of mechanical and electrical components in shielded rooms, without the need for additional back-up shielding behind devices and components. Another object in at least an embodiment is to allow for easy renovation, removal, reinstallation and recycling of the components of the system. Another object in at least an embodiment is to make the system fit (match) pre-existing standard components (re: 6″ metal standard track). Another object in at least an embodiment is to make repairs to drywall on the wall system easier by preventing damage to the radiation shielding lead sheet material by placing it in or near the centre of the wall cavity. Another object in at least an embodiment is to protect the health of construction workers renovating or removing such walls by suspending lead at the centre of the wall cavity and by making the lead easy to remove, store or reuse. Also, each lead panel may be painted both sides to protect those who handle these panels and marked as to its Pb content. Therefore, there is provided in one or more embodiments a radiation shielding leak-proof barrier system having a new metal stud configuration that permits elimination of the need for use of lead laminated gypsum board, elimination of the need for use of lead screw caps, and elimination of the need for use of lead barrier tabs/overlap plates. Some embodiments permit elimination of the need for the installers to simultaneously handle both the drywall and lead sheets and thus reduces the chance of injury to the installer by reducing material weight during construction and thereby reduces both the quantity and severity of injury claims by installers. In one or more embodiments, the system assists in preventing damage to the lead and thus compromise to the integrity of the shielding in the event of an impact with the drywall either during construction or during normal use of the completed building. In one or more embodiments, the system assists in preventing future sag and deformity of lead sheets by method of continuous support bars acting as continuous support for the lead sheet. In one or more embodiments, the system assists in ensuring the wall will not fail due to incorrectly sized metal studs being used as only the specific studs as designed will accept the support bar. In one or more embodiments, the system assists in reducing construction time by way of the components being prefabricated into simple and readily assembled components. In one or more embodiments, the system assists in providing maximum protection against radiation leakage by way of sheets of radiation shielding material being properly overlapped, supported continuously along their length, and installed to the required height and in the required positioning due to the design of each component and the method of assembly of the complete radiation shielding barrier system. Radiation shielding material may comprise lead or any other suitable radiation shielding material. The type of material depends on the radiation intended to be attenuated by the material. Thus, for gamma/X-rays emitted for example by an X-ray source in a room defined by the disclosed walls, a suitable material would be a dense metal such as lead. Wall elements may include studs and support bars, not necessarily exactly as disclosed herein, but suitable for supporting sheets of radiation shielding material. A metal stud in an embodiment is configured to accept metal support bars and metal restraining bars, configured to accept radiation shielding lead panels, and assembled so that when assembled according to provided instructions, radiation shielding lead sheets will overlap and thereby create a radiation leak proof metal stud system. Leak proof in this context means that there is no free path for radiation to pass through the wall without encountering radiation shielding material. In an embodiment in which the radiation shielding lead panels are suspended and fastened from metal support bars, no penetrations need be made in the portion of the lead that faces the room, enabling the creation of a radiation leak proof metal stud system. Furthermore, radiation shielding lead panels may be sized appropriately according to the structural strength of the metal support bars and held and secured to the metal support bars, the radiation shielding lead panels may be evenly supported and no portion of the radiation shielding lead panel need be stressed from a single fastener, and thus the radiation shielding lead panels may remain in place without deforming for the lifespan of the building. Furthermore, the use of radiation shielding lead panels do not need any drywall screws to penetrate the lead and this eliminates the need for lead screw caps, and thereby enabling the creation of a radiation shielding leak proof metal stud system. Mechanical, electrical or any other services would be installed on the side of the radiation shielding barrier, it is to serve, to avoid penetrating the radiation shielding barrier. Furthermore, in at least an embodiment, electrical devices, such as for example, switches or receptacles; or plumbing fixtures, such as water supply and drainage lines, are able to be installed as they would be in a traditional metal stud wall, without damage to the radiation shielding barrier and without the use of additional shielding, thereby simplifying the construction process and maintaining a radiation leak proof metal stud system. Should additional room be needed for the placement of electrical or mechanical items either side of the radiation shielding barrier, the installation of a furring wall could be placed of a thickness required. Furthermore, during renovations or removal of existing lead lined gypsum board systems, precautions must be taken to prevent contamination of the work area and danger to the health of personnel due to the fact that lead is a toxic material. If renovations are required to walls consisting of this metal stud system, such renovations may be more easily accomplished as the gypsum board need not itself contain lead, the radiation shielding lead material panels are painted to encapsulate the lead and may be easily removable and in at least an embodiment may be stored, reused, or easily recycled. Furthermore, if repairs to the drywall surface are required, in at least an embodiment such repairs can be completed without the need to repair the radiation shielding lead surface. In an embodiment there is provided a metal stud system of accepting and supporting radiation shielding lead panels for radiation shielding wall construction, wherein the metal stud is configured with an ‘s’ bend in it to ensure proper placement and alignment and lead overlap, and is configured to accept metal support bars which fit into a specific configuration to allow for proper radiation shielding lead support and alignment and overlap, which in turn support preformed radiation shielding lead panels that are configured to fit into the wall system and provide for proper support, fastening and overlap. Although in this embodiment radiation shielding lead panels are used as sheets of radiation shielding material, other embodiments may use any other suitable radiation shielding material. FIG. 1 shows a perspective view of an assembly comprising a top track 18, bottom track 16, five studs 14, and twelve radiation shielding lead panels 12. Additionally, twelve metal support bars and four metal restraining bars would be present in this assembly but are not visible in this illustration. The radiation shielding lead panels 12 can be a variety of thicknesses to suit the intended application. Typical installations (suitable for most hospital diagnostic imaging rooms) will use 4 lbs/sq ft (1.58 mm thick) lead, 6 lbs/sq ft (2.37 mm thick) lead and 8 lbs/sf (3.16 mm thick) lead. The radiation shielding lead panel may be made of sheet lead, configured to a specific shape and bent 90 degrees at fold line to create a foldover edge (22) (not shown in FIG. 1). The studs 14 may be made of galvanized sheet steel, thickness 20 gauge but can be made in different thicknesses if required. The studs must be made of steel or similar high strength material. The studs are currently available 10′ long but can be any length to suit project requirements. The studs may have a dimpled texture pressed into the surface of the two legs that are in contact with drywall. This feature is not present in, but could be added to, the embodiment shown in the figures. This is to provide better grip for drywall screws when they are being installed. The top track 18 and bottom track 16 are pre-manufactured items. The top track can be a single or double top track or in another configuration if required. FIG. 2 shows a side section view through the assembly of FIG. 1, showing radiation shielding lead panels 12, metal support bars 20, a stud 14, top tracks 18, and bottom tracks 16. In the embodiment shown inner top track 18 is set into but not fastened to outer top track 18a, an arrangement known as deflection track assembly, which allows for normal movement of the structure without causing drywall cracking in the wall below. Lead shielding panels 12 are hung from support bars 20 which are suspended between studs 14 (only one stud is visible in this view). The holding or support bars 20 span between two studs 14 and attach to the studs in predetermined locations (pre-drilled, or with holes formed by punching or in some other fashion) 40 with fasteners 26 (pre-drilled or punched locations and fasteners not shown in FIG. 2). The bars 20 are designed to accept radiation shielding lead panels 12 so that foldover edge 22 of radiation shielding lead panel 12 folds over the top of the support bar and is fastened with mechanical fasteners 24 (not shown in FIG. 2). The bars 20 are U-shaped in cross section with two tabs 42 (one on each end) sticking upwards to allow fastening. The tabs have in this embodiment slots 44 for receiving fasteners to attach the tabs to the studs. The bars may be made of galvanized sheet metal, 20 gauge or to suit project requirements. The foldover edge 22 of radiation shielding lead panel 12 sits on top of metal support bar 20. The installer should secure radiation shielding lead panel 12 through the foldover edge 22 with mechanical fasteners 24. The mechanical fasteners help prevent the lead radiation shielding panel from falling off if the wall is disturbed or if the panel is bumped during other construction work. FIG. 3 shows a top section view through a portion of the assembly of FIG. 1, showing a stud 14, support bars 20, radiation shielding lead panels 12, and bottom tracks 16. As can be seen in FIG. 3, the stud 14 has a shape having legs 14a, here 2.5 inches long but can be adjusted as required, offset leg 14b, 1 inch long to allow overlap as prescribed by code and not recommended to have any other length, and leg 14c, here 3.5 inches long but can be adjusted as required. The stud 14 is elongated in the direction perpendicular to the cross-section. The stud also has legs 14d and 14e at the ends of the cross-section visible in FIG. 3. Attached to stud 14 by mechanical fastener 46 is restraining bar 28. The restraining bar 28 may be an L-shaped clip, or may be another shape as required. This component may not be necessary, i.e. it is only needed if the radiation shielding lead is bent (ie old panel being reinstalled, if the wall is sloped to the extent that the radiation shielding lead hangs away from the surface of the stud (14), drafts or air movement in the wall necessitate additional restraining of the radiation shielding lead, or other reasons why the radiation shielding lead may not sit flat). The restraining bar 28 may be made of metal or similar rigid material; in the embodiment shown it is made of 20 gauge galvanized steel. A further mechanical fastener 26 attaches support bars 20 to stud 14. This fastener is currently shown as a small bolt & nut. It is preferably a bolt & nut or similar fastener as this will use preexisting holes 40 punched or drilled in the metal stud leg 14a and slot 44 in the support bar 20. Although a single fastener is shown to attach two support bars to a stud, in alternate embodiments multiple fasteners could be used for each support bar and each fastener may be used to attach only a single support bar to a stud. For example, the support bars could have tabs with multiple prongs that are not aligned with the prongs on the corresponding support bar on the opposite side of a stud, and each prong having a fastener, there being a corresponding hole in the stud for each fastener. FIG. 4 shows a section view facing forwards through the connection point from the stud 14 to two metal support bars 20 in the assembly of FIG. 3. In this figure are shown the stud 14, support bars 20 with tabs connected to leg 14a of the stud 14 with fastener 26, and foldover edge 22 of radiation shielding lead sheets 12 lying on top of support bars 20 (main portion of radiation shielding lead sheets 12 not visible in the figure as they lie behind the plane of view). FIG. 5 shows a section view facing rearwards through the connection point from the stud 14 to two metal support bars 20 in the assembly of FIG. 3. In this figure are shown the stud 14, support bars 20 with tabs connected to leg 14a of the stud with fastener 26, and foldover edge 22 of radiation shielding lead sheets 12 lying on top of support bars 20. Foldover edges 22 of radiation shielding lead sheets 12 are attached to support bars 20 with mechanical fasteners 24. These are currently shown as self-tapping screws, but could be regular screws, rivets, crimps, pins, clamps, or some other form of mechanical fastening device. In this view the radiation shielding lead sheet 12 on the right hand side of the figure extends behind leg 14b of stud 14 to overlap with the radiation shielding lead sheet 12 on the left hand side of the figure. FIG. 6 shows a way to use the metal stud in the present invention to create a radiation shielding leak-proof shielding barrier around electrical and mechanical equipment and receptacles without the need for special or custom construction, while maintaining the integrity of the shielding. Finishing material 34 is attached to the outside edges of stud 14 to cover the sides of the wall. Finishing material 34 may be gypsum board or any other type of wall finishing product or material. A finish material acts to protect the radiation shielding lead from damage. Pipe 32 may also be a chase, conduit, or other mechanical or electrical or other piece of equipment or hardware typically concealed inside a wall. Electrical equipment 36 may also be another electrical or mechanical equipment or device or control. Pipe 32 and electrical equipment 36 are shown to indicate that equipment such as this can be installed in the wall without the need for additional radiation shielding. FIG. 7 shows a way to use the metal stud at a corner where a radiation shielded wall meets an unshielded radiation shielding wall, while maintaining the integrity of the radiation shielding. A metal corner assembly 30 may comprise radiation shielding lead laminated to sheet metal. The corner assembly as shown in FIG. 7 is for a bend in the shielded wall, which could be unconnected to any unshielded walls or at a T-junction or + junction. Variants of the corner assembly could be used at a T-junction or a + junction between any combination of shielded or unshielded wall assemblies. The legs can be of varying length to suit project requirements, and the angle of the centre bend (shown 90 degrees) can be any angle or combination of angles, including 0°, 180°, 360° etc., as required to suit project requirements. The corner assembly 30 may be fastened to stud 14 with mechanical fasteners 26. Also visible in this figure is commercially available metal stud 38, showing how current metal stud 14 is the same overall width and depth as a commercially available metal stud 38 and thus the current metal stud 14 can be used with standard components from commercially available metal stud wall systems, and showing how radiation shielded and unshielded walls can be constructed together easily. FIG. 8 illustrates a radiation shielding lead panel 12, not to scale, shown with the fold line between the main body of radiation shielding lead panel 12 and foldover edge 22 of radiation shielding lead panel 12 indicated by a solid line. Lead panels are sized to meet requirements of the International Lead Association (lead). The lead panel may be color-coded to indicate the thickness of the panel, for example by painting some or all of the surface of the lead panel. Horizontal hatching in FIG. 8 indicates a color. FIG. 9 illustrates a different lead panel with a different color indicated by vertical hatching. FIG. 10 is a plan detail of another embodiment of a radiation shielding panel, shown at a sheet metal fastening bracket x1. The bracket is attached to a substrate x2, such as ¼″ plywood, although it could be a different thickness or material. A lead sheet x3 of any required thickness is laminated to the substrate. The lead sheet may extend a length x4, for example ⅞″, beyond one end of the bracket. Mechanical fasteners 24, which may be screws or another fastener such as a rivet, attach the fastening bracket to the substrate but preferably do not pierce the lead. FIG. 11 is an elevation detail of the panel of FIG. 10. Wavy lines in FIG. 11 indicate continuing surfaces. As can be seen in FIG. 11, the sheet metal fastening bracket x1 extends across the panel but covers only a portion of its surface area. FIG. 12 is an overall elevation of the panel of FIG. 11, showing several fastening brackets x1 traversing the panel and a strip of lead x4 extending from one edge of the panel. FIG. 13 is a section detail of the panel of FIG. 12 showing an end return x5 of a metal fastening bracket x1. The end return may be 1″ square and may have a slotted hole punched in it and in the corresponding return at the other end of the bracket. The panel of FIGS. 10-13 may be color-coded to indicate the thickness of the lead, for example by painting some or all of the surface of the lead. FIG. 14 is a plan detail of two panels as shown in FIGS. 10-13 meeting at a stud. This view is similar to that shown in FIG. 3 but with different panels. FIG. 15 is a plan detail of a deep enclosure box x8. The deep enclosure box is a sheetmetal box designed so that you could recess a deep item (such as an electrical panel) inside the shielded wall, without affecting shielding integrity. It uses the same typical metal stud and is 2100 MM tall, or to suit the job site requirements. The standard width is sized to suit 16″ o.c. studs however this can also be sized to suit the job site requirements. A piece of lead x7 is laminated to one side of the deep enclosure box to provide radiation protection. Another piece of lead x6 is attached to an end of the deep enclosure box to provide an overlap with another panel (not shown) that would extend across the other side of offset leg 14b. Piece of lead x6 may also be attached by lamination. FIG. 16 is an overall plan of the deep enclosure box of FIG. 15. FIG. 17 is an elevation view of a wall portion, showing four studs 14, six typical panels 12 in the two side stud cavities, and a deep enclosure box x7 visible in the middle. Immaterial modifications may be made to the embodiments described here without departing from what is covered by the claims. In the claims, the word “comprising” is used in its inclusive sense and does not exclude other elements being present. The indefinite articles “a” and “an” before a claim feature do not exclude more than one of the feature being present. Each one of the individual features described here may be used in one or more embodiments and is not, by virtue only of being described here, to be construed as essential to all embodiments as defined by the claims. |
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abstract | Systems and methods for left radial access, right room operation peripheral interventions are provided that include left radial bases to stabilize a left arm of a cardiac patient across a midsagittal plane, transradiant right radial bases to position a right arm of the patient, and radiodense radiation reduction barriers located between the patient and a doctor. |
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description | The present application claims priority from Japanese application JP 2005-174491 filed on Jun. 15, 2005, the content of which is hereby incorporated by reference into this application. The present invention relates to a method for inspecting electrical defects of a fine circuit formed on a semiconductor wafer and an apparatus thereof. As a method of detecting the defects of a circuit pattern formed on a wafer by comparative inspection of an image in a production process of a semiconductor device, JP-A No. 258703/1993, for example, describes a method of comparatively inspecting a pattern by a so-called SEM method wherein an electron beam focused into a spot is used for scanning. The feature of a SEM inspection apparatus is that the resolution thereof is higher than that of an optical inspection apparatus and it can detect electrical defects. However, since a SEM inspection apparatus is based on a method of focusing an electron beam into a spot and obtaining an image by two-dimensionally scanning the surface of a specimen, when a specimen is inspected with the apparatus, long scanning time is required and thus the apparatus has an essential drawback to the future increase of inspection speed. Further, as an electron beam inspection method attempting to obtain a higher speed, JP-A No. 249393/1995, for example, describes an inspection apparatus of a projection type wherein a semiconductor wafer is irradiated with a rectangular electron beam and generated reflection electrons and secondary electrons are focused into an image with an electron lens. A projection type inspection apparatus can be expected to form an image at a higher speed than the SEM method since it can irradiate an object at a time with an electron beam of a higher current than the SEM method and can obtain an image in an integrated manner. However, in the case of a projection type inspection apparatus, a problem is that the distribution of the emission angle of secondary electrons is wide at the time of imaging. The distribution of the emission angle of secondary electrons follows the cosine rule and hence most of the secondary electrons are emitted at a large angle on the basis of the direction of the normal to a wafer. When all of such secondary electrons are taken into an objective lens and focused into an image, a sufficient spatial resolution cannot be obtained due to the aberration of the objective lens. In order to obtain a sufficient spatial resolution of a 100 nm level, it is necessary to form an image while the secondary electrons used are limited to those emitted at angles within a small angle of aperture (0.1 rad for example) to the axial direction of a lens. Therefore, even though a high current electron beam is used for irradiation as an areal beam in order to form an image, the proportion of the secondary electrons capable of actually contributing to the imaging is low and hence a required S/N ratio of an image is hardly obtained. In the case of using reflection electrons too, the obtained emission amount is smaller by double digit in comparison with the electric current of the irradiation beam and it is difficult to obtain both high defect detection sensitivity and high speed inspection simultaneously with a conventional projection type inspection apparatus. In the meantime, as a method of securing both high sensitivity defect detection and high speed detection, JP-A No. 108864/1999 discloses a projection type wafer inspection apparatus wherein electrons pulled back before they impinge with a specimen by field reversing immediately above a wafer (hereunder referred to as mirror electrons or mirror reflection electrons) are used as imaging electrons. A mirror electron imaging type inspection apparatus has two main features which are different from the features of a conventional projection type inspection apparatus wherein secondary electrons and reflection electrons are focused into an image. The first feature thereof is that mirror electrons from a specimen do not have such a wide angle distribution as secondary electrons have and are emitted nearly straight above the surface of a specimen, and hence it is possible to take almost all of the electrons into an imaging lens system and increase the amount of image signals. The second feature thereof is that, in a region where incoming electrons are mirror-reflected immediately above a specimen, the kinetic energy of the electrons reduces considerably and the track changes in accordance with even a slight deviation of a surface, and hence the difference in image contrast between defective portions and normal portions increases. It means that the load for image processing reduces to the extent that the difference in image contrast between defective portions and normal portions increases in comparison with a secondary electron and reflection electron imaging type inspection apparatus that obtains an image of a high resolution and detects slight difference of the image. In addition to those features, in a mirror electron imaging type inspection apparatus, most of the irradiation electrons are reflected immediately above a wafer and hence basically they do not enter into the wafer. Electrons having slightly higher energy exist in an electron beam since the electron beam has an energy distribution and those electrons enter passing through a potential barrier. However, the value is several eV at most. That is, a mirror electron imaging type inspection apparatus can deal with even a specimen which has the fear of damage caused by an electron beam with a SEM inspection apparatus or a secondary electron imaging type inspection apparatus as an object of the inspection. A mirror electron imaging type inspection apparatus sensitively detects potential change formed by the unevenness of a surface and can have good sensitivity also to electrical defects formed on a wafer in the same way as a SEM inspection apparatus. For example, when a defect of no electrical-conductivity exists, since the portion is electrically insulated, the electric potential on the surface of the portion can be differentiated from that of an electrically conductive normal portion by electrification, and the abnormality of the potential can be detected by using mirror electron imaging. However, in a mirror electron imaging type inspection apparatus, an irradiation electron beam is mostly repulsed immediately in front of a wafer by field reversing and hence it is impossible to control the electrification of a specimen with the irradiation electron beam. As a consequence, it becomes necessary to control the state of the electrification of the surface of a specimen before the irradiation of a primary electron beam (preliminary electrification) in order to obtain a stable inspection image. The preliminary electrification can be carried out by: a method of irradiating a specimen to be inspected with light including ultraviolet rays or an electron beam having energy enough to generate secondary electrons; a method of applying a prescribed potential to the surface of a specimen; or another method. JP-A No. 14485/2004 describes a preliminary electrifier to electrify a wafer before inspection. The electrification potential formed on the surface of a wafer by applying preliminary electrification varies in accordance with the type of an insulator and a circuit pattern and, since electric charge escapes little by little, electrification potential decreases at a certain time constant. Such a time constant is sufficiently long in comparison with the time required for obtaining a mirror electron image but is insufficient in comparison with the inspection time of a whole wafer, and thus additional preliminary electrification is required during inspection. In order to detect electrical defects and therefor control the electrification potential of a wafer by the preliminary irradiation of an electron beam, a control electrode is disposed immediately above the wafer. The control electrode in the case of JP-A No. 14485/2004 is configured so as to transmit an irradiation electron beam and apply electric potential immediately onto the wafer by using a grid-shaped electrode. The principle of the control of the electrification potential by a grid electrode is explained hereunder. When preliminary irradiation is applied, the value of the irradiation energy of an electron beam is set beforehand so that the secondary electron emission efficiency may be one or more. In the case of a general insulative film material for a semiconductor device, the value is about 500 V. The surface of an insulative film formed on a wafer is positively electrified gradually by the irradiation of an electron beam since the secondary electron emission efficiency thereof is larger than one. When a potential relatively positive to the potential of a wafer surface is applied to a control electrode, the generated secondary electrons are pulled toward the control electrode and hence the wafer surface is positively electrified gradually. When the electrification potential of the wafer surface equals to the potential of the control electrode, then the electric potential gradient between the control electrode and the wafer surface is leveled and hence the generated secondary electrons begin to return to the wafer surface. As a result, the positive electrification of the wafer surface is alleviated, the electric potential gradient between the control grid and the wafer surface reappears, and the secondary electrons are pulled toward the control electrode again. As a consequence, the electrification potential of the wafer surface balances with the potential of the control electrode at a nearly equal potential level. When a potential relatively negative to the potential of a wafer surface is applied to a control electrode inversely, the generated secondary electrons are pushed back from the control electrode and return to the wafer surface and hence the effective secondary electron emission efficiency becomes lower than one. In consequence, the wafer surface is negatively electrified until the electric potential gradient between the control electrode and the wafer surface is leveled. By so doing, the electrification potential of a wafer surface is controlled with a control electrode. When it is attempted to further increase inspection speed and improve defect detection accuracy in the inspection of a wafer pattern, conventional technologies have had the following problems. In the case of a mirror electron imaging type inspection apparatus, when electrical defects in a wafer are electrified by preliminary irradiation, it is necessary that the electrification potential is uniform in the entire preliminary irradiation region. The reason is that, in a mirror electron imaging type inspection apparatus, since an image is formed by using the reflection of irradiation electrons at a certain potential plane of a wafer surface, if the electrification potential of the wafer varies even slightly, the distance between the potential plane at which the irradiation electron beam is reflected and the wafer surface also varies and, as a result, the imaging conditions vary and the contrast of the mirror electron image also varies. As a result of the present inventors' experiments, it has been found that the allowable variation of the electrification potential is about 0.5 V or less and such a uniform electrification potential distribution cannot be attained only by simply disposing a grid electrode and applying electron beam irradiation. Further, there are some cases where a preliminary electrifier is applied also to a conventional SEM inspection apparatus. However, in the case of a SEM inspection apparatus, secondary electrons generated when irradiation electrons enter a wafer during inspection electrify the irradiation region again and the detection efficiency of the secondary electrons does not vary with the variation of specimen potential being about several volts, and hence the highly accurate uniformity by the preliminary electrification is scarcely required. Here, in the case of a mirror electron imaging type inspection apparatus, most part of the irradiation electron beam is repulsed immediately in front of a wafer by field reversing and hence it is impossible to control the electrification of a specimen with the irradiation electron beam. Furthermore, in the case of the mirror electron imaging method, the required uniformity of the electrification of a specimen surface is as strict as in the range of about plus or minus 0.5 V and sufficient uniformity has not been obtained with a conventional preliminary electrification technology used for a SEM method. When a conventional preliminary electrification technology is applied to a mirror electron imaging type inspection apparatus, concretely it is estimated that the following problems arise. In an electrification controller, a problem is that a grid electrode is used as the control electrode and the distribution of electrification potential becomes two-dimensionally uneven. In the case of applying electrification control by the irradiation of ultraviolet rays or an electron beam through a grid-shaped electrode, the irradiation is not applied to the portions of a wafer corresponding to the grid, and hence the supply of electric charge is insufficient and the portions where an intended electrification potential is not attained are undesirably formed. Electrification control by preliminary irradiation can be carried out at any time including before or during the inspection of a specimen to be inspected, but it goes without saying that such unevenness of electrification badly affects the inspection accuracy. When preliminary irradiation is applied during inspection with a preliminary electrifier or the like, the unevenness of electrification increases particularly in the direction perpendicular to the wafer moving direction. A wafer moves relatively to a preliminary electrifier and the irradiation regions partially overlap with each other in the wafer moving direction in many cases. Therefore, the electrification state is equalized in the direction parallel with the moving direction and the unevenness is alleviated to some extent. However, with regard to the unevenness in the direction perpendicular to the moving direction, such equalization is not applied and hence the unevenness in electrification potential still remains. As stated above, a problem has been that an abnormal contrast caused by the unevenness of electrification potential in a mirror electron image is falsely counted as a defect in real inspection and correct inspection is hindered. The present invention is characterized by adopting a means of always uniformly electrifying a desired region including the inspection region of a specimen before obtaining an inspection image by mirror electrons. More specifically for example, the present invention is configured so that the electrification potential of a specimen may be equalized along with the movement of a stage by: disposing preliminary electrifiers onto a mirror electron imaging type inspection apparatus; then disposing not a grid-shaped electrode but an electrode having a slit-shaped opening as an electrode to control the electrification potential of each preliminary electrifier; and directing the longitudinal direction of the opening to the direction perpendicular to the moving direction of the stage. Further, the present invention is configured so that: the potential gradient at the boundary of the preliminary electrification region may be reduced by decreasing the irradiation strength as the measurement point comes close to the boundary of the preliminary electrification region; and thus sufficient equalization may be obtained by additional preliminary electrification. The present invention makes it possible to: form always uniform electrification potential on a wafer; and thereby detect defects of a semiconductor pattern without error at a high speed. Configurations of examples according to the present invention are hereunder explained in reference to drawings. FIG. 1 shows the outline of a mirror electron imaging type inspection apparatus on which preliminary electrifiers are mounted. Note that a vacuum pump for evacuation, a controller thereof, exhaust pipes, and others are not shown in the figure. Firstly, major elements of an electron optical system of the present apparatus are explained. An irradiation electron beam 100a emitted from an electron gun 101: is deflected with an ExB deflector 103 while being converged with a condenser lens 102; forms a crossover 100b; and thereafter is emitted in the form of nearly parallel rays onto a specimen wafer 104. Although the condenser lens 102 is expressed as single lens in the figure, it may also be a system combining plural lenses in order to optimize the optical conditions. In the present embodiment, a Schottky electron source of a Zr—O—W type is used as the electron gun 101. An electron gun using a Zr—O—W type Schottky electron source is suitable for the present apparatus which is aimed at high speed inspection since it can stably supply a uniform electron beam having a large electric current (1.5 μA for example) and an energy width of 1.5 eV or less. In the present invention however, the electron source is not limited to a Zr—O—W type electron source and any electron source is useful as long as it is an electron source which is small in size and has high luminance, and thus an electron source which uses carbon nanotubes or the like can also be used for example. The voltage and current necessary for operation such as extraction voltage to the electron gun 101, acceleration voltage to the extracted electron beam, heating current to an electron source filament, and others are supplied and controlled with an electron gun controller 105. The ExB deflector 103 is placed in the vicinity of an imaging plane 100d on which an imaging electron beam 100c is focused. In this case, the aberration of the irradiation electron beam 10a is generated by the ExB deflector 103. When the correction of the aberration is required, another ExB deflector 106 for aberration correction is placed between the irradiation condenser lens 102 and the ExB deflector 103. The irradiation electron beam 100a deflected by the ExB deflector 103 so as to be parallel with the axis perpendicular to the wafer 104 is formed with an objective lens 107 into an areal electron beam entering in the direction perpendicular to the surface of the specimen wafer 104. A fine crossover is formed on the focal plane of the objective lens 107 with the irradiation condenser lens 102 and hence it is possible to irradiate the specimen wafer 104 with an electron beam having good parallelism. The region of the specimen wafer 104 which is irradiated with the irradiation electron beam 100a is set so as to have a large area; for example, 2,500 μm2 or 10,000 μm2. A negative voltage nearly equal to or slightly lower than (larger than in absolute value) the acceleration voltage of an electron beam is applied to the specimen wafer 104 mounted on a wafer stage 108. By this negative voltage, the irradiation electron beam 100a: scarcely impinges with the wafer 104; is decelerated short of the wafer and reflected; and is pulled back upward as mirror electrons. The supply and control of the voltage applied to the wafer 104 are carried out with a wafer voltage controller 109. It is necessary to adjust the difference from the acceleration voltage of the irradiation electron beam 100a with a high degree of accuracy in order to reflect the irradiation electrons in the very vicinity of the wafer, and thus the control is carried out in conjunction with the electron gun controller 105. The electrons coming flying from the side of the wafer: reflect information related to the electrical defects of a circuit pattern on the wafer 104; and are taken into the apparatus as an image for the defect judgment by imaging using an electron imaging optical system. The mirror electrons: are subjected to focusing effect with the objective lens 107; move upward as they are in the vertical direction since the ExB deflector 103 is controlled so as not to apply deflection effect to the electron beam coming from the lower side; and are projected onto an image detector 112 in an expanded manner with an intermediate lens 110 and a projector lens 111. The projector lens 111 is described as single lens in the figure, but it may be composed of plural electron lenses for the purpose of securing a high magnification and correcting image distortion. The image detector 112 transforms an image into electric signals and sends the distribution of local electrification potentials on the surface of the wafer 104, namely a defect image, to an image processor 116. In the electron optical system, in addition to the units explained here, auxiliary electron optical devices such as an aligner to correctly transfer an electron beam to each electron optical device, a stigma to correct the distortion of an image, and others are disposed in an appropriate manner. However, those are omitted in the figure. The control of an electron optical system including such auxiliary electron optical devices is implemented with an electron optical system controller 113. Next, the image detector 112 is explained. The detection of an image is carried out by optically coupling a fluorescent screen 112a to transform a mirror electron image into an optical image with an optical image detector 112c through an optical image transfer system 112b. In the present embodiment, an optical fiber bundle is used as the optical image transfer system 112b. The optical fiber bundle is formed by bundling fine optical fiber elements of the same number as the number of pixels and can efficiently transfer an optical image. Further, when a fluorescent image having a sufficient amount of light is obtained, the optical transfer efficiency may be lowered, and thus it is also possible to: use an optical lens in replace of the optical fiber bundle; and focus an optical image on the fluorescent screen 112a onto the light receiving surface of the optical image detection element 112c with the optical lens. In this case, the flexibility of the optical transfer system increases and image processing such as the expansion and distortion correction of an optical image and the like can be carried out more easily. Furthermore, it is also acceptable to: insert an amplifier into the optical image transfer system; and make it possible to transfer an optical image having a sufficient amount of light to the optical image detector 112c. The optical image detector 112c transforms an optical image focused on the light receiving surface into electrical image signals and outputs the signals. As the optical image detector 112c, a CCD, an MCP (micro channel plate), a photodiode, or the like can be used. Otherwise, a TDI sensor using a time accumulation type CCD may be used. The wafer stage 108 must be driven in synchronization with the timing of image capture at the image detector 112. In the case where a TDI sensor is used as the optical image detector 112c in particular, the image resolution considerably lowers unless the transfer of image signals among pixels is correctly synchronized with the movement of the wafer stage 108. For that reason, the wafer stage is equipped with a position sensor 114 to detect the position of the wafer stage, the drive of the stage is controlled with a stage controller 115, and the timing of image capture is transmitted to the image detector 112 on the basis of the information on the stage position coming from the position sensor 114. The image processor 116 comprises an image signal storing unit 116a and a defect judgment unit 116b. The image signal storing unit 116a: obtains electron optical conditions, image data, and stage position data from the electron optical system controller 113, the image detector 112, and the stage controller 115, respectively; and stores image data in the manner of being liked to a coordinate system on the specimen wafer. The defect judgment unit 116b: uses the image data linked to the coordinate on the wafer; and judges a defect by various defect judging methods of comparing the image data with a prescribed value, an adjacent pattern image, an image at the same pattern position in an adjacent die, and the like. The coordinate of a defect and the signal strength of corresponding pixels are transferred to and stored in an inspection apparatus controller 117. Such a defect judging method is: either set by a user or coordinated with the type of a wafer beforehand; and selected by the inspection apparatus controller 117. Operation conditions of each section of the apparatus are input from the inspection apparatus controller 117. Into the inspection apparatus controller 117, various conditions such as an acceleration voltage at the time of electron beam generation, an electron beam deflection width and deflection speed, a wafer stage traveling speed, the timing of image signal capture from image detection elements, and others are input beforehand. Then the inspection apparatus controller 117 controls the controllers of various elements in an integrated manner and acts as an interface with a user. The inspection apparatus controller 117 may also comprise plural computers which share the roles and are connected through communication circuits. Further, a monitor-mounted I/O device 118 is installed. With the mirror electron imaging type inspection apparatus of the present embodiment, a specimen wafer is electrified insufficiently since the electron beam scarcely impinges with the wafer 104. However, it is necessary to sufficiently electrify a specimen wafer to the extent of creating a difference from normal portions in order to detect an electrical defect. For that reason, preliminary electrifiers 119 are provided. Both preliminary electrifiers 119a and 119b are controlled with a preliminary electrification controller 120. The preliminary electrification controller 120, in conjunction with the wafer voltage controller 109 and the electron gun controller 105, controls the electrification potential on a wafer formed with the preliminary electrifiers 119a and 119b so as not to disturb the state wherein an irradiation electron beam is reflected in the very vicinity of a wafer surface. The details of the operations are explained with FIG. 2. In FIG. 2, an evacuating system to keep the apparatus in the state of vacuum, an image processing system, a stage control system, an electron optical system, a control system, and others are not shown. FIG. 2 is a view of a mirror electron imaging type inspection apparatus viewed from above. In FIG. 2, the details of an electron optical system column are omitted and only the position of an objective lens 107 for mirror electron imaging is shown. A wafer stage 108 held in a vacuum chamber 201 moves little by little in the Y direction while moving reciprocally in the X direction as shown with the arrows in the figure so as to be able to inspect the entire wafer. Each of preliminary electrifiers 119a and 119b is allocated on each side of the objective lens 107. The object thereof is: to efficiently apply preliminary electrification to a wafer moving reciprocally to the column immediately before inspection; and also to electrify the wafer before the inspection and eliminate the electrification after the inspection. The control system and electric wiring of the preliminary electrifiers 119a and 119b are omitted in the figure. The wafer 104 to be inspected is firstly contained in a wafer cassette or the like and set at a wafer charging port 203. Thereafter the wafer is transferred into a preliminary chamber 202 open to the air with a transfer robot 204. Successively a charging port of the preliminary chamber 202 is closed, thereafter the preliminary chamber 202 is evacuated to form a vacuum with a vacuum pump, and the wafer 104 is transferred into the specimen chamber 201 which is always maintained in the state of vacuum without deteriorating the vacuum in the specimen chamber 201. When the wafer is transferred, the wafer stage 108 moves to the position shown by the reference character HP in the figure and receives the wafer. After the wafer is transferred into the specimen chamber 201, a gate to and from the preliminary chamber 202 is closed. When a TDI sensor is used for the optical image detector 112c, the direction of integrating pixel signals conforms to the direction of continuous movement of the wafer stage 108. Further it is also necessary for the preliminary electrifier to be located in parallel with the direction of the continuous movement of the wafer stage when preliminary electrification is applied during inspection. In the inspection of a wafer, there are some cases where the direction of the continuous movement of the stage is rotated by 90 degrees in accordance with the arrangement of the pattern of the wafer to be inspected. That is, although, when a pattern to be inspected is arranged densely in the X direction in the figure and sparsely in the Y direction therein, the stage movement shown by the arrow A in the figure is desirable, when a pattern is arranged in the opposite manner, it is desirable to rotate the directions by 90 degrees. However, the attaching angle of the TDI sensor is constant usually, namely the integral direction of pixel signals is the X direction in the case of FIG. 2, the preliminary electrifiers 119 are also arranged in the X direction, and, if nothing is done, it is impossible to rotate the movement direction for inspection by 90 degrees as stated above. Therefore, the transfer robot 204 is provided with mechanism wherein the wafer set at the wafer charging port 203 is rotated in the movement direction for inspection assigned by a user when the wafer is transferred to the preliminary chamber 202, and the wafer is placed on the wafer stage of the preliminary chamber 202. In addition to the mechanism, it is also possible to: dispose two types of wafer charging ports 203; and use a type thereof when the inspection movement shown by the arrow A is underway and the other type thereof when the inspection movement in the direction rotated by 90 degrees is underway. Otherwise, although the transfer robot 204 transports a wafer to the preliminary chamber 202 while the direction of the wafer is always the same, it is also possible to: rotate the wafer stage of the preliminary chamber 202 in conformity with the inspection movement assigned by a user after the wafer is transported; and then transfer the wafer to the vacuum chamber 201 so as not to create contradiction between the direction of the wafer setting and the direction of the continuous movement of the wafer stage. Still otherwise, it is also possible to: add a wafer rotating mechanism to the wafer stage 108 itself; and satisfy the relationship predetermined by a user between the direction of the wafer setting and the direction of the continuous movement of the wafer stage in the vacuum chamber 201. The inspection operations after the placement of a wafer is finished are explained taking the case where the inspection is carried out from the right upper portion of the wafer in FIG. 2 as an example. Firstly, the wafer stage 108 moves to the position where the first inspection region is located on the left side of the preliminary electrifier 119a. When the inspection starts, the wafer stage 108 moves toward the right side, the preliminary electrifier 119a carries out preliminary irradiation under predetermined conditions, and thus the wafer is electrified sequentially. When the wafer passes through the objective lens 107, a mirror electron image is obtained and a defect is detected from the inspection image. Thereafter, electrification is applied again if necessary when the wafer passes through under the preliminary electrifier 119b. The additional electrification includes the elimination of the electrification when the electrification potential is set at 0 V. After the inspection region passes through the preliminary electrifier 119b, the wafer slightly moves upward in the figure by a distance predetermined in the Y direction, and moves to the reverse side in the X direction. Since the direction of the wafer movement is reversed, in this scanning, the preliminary electrifier 119b electrifies the wafer up to an electrification potential required for the acquisition of a mirror electron image and the preliminary electrifier 119a plays the role of the additional electrification. The switching of the conditions of the preliminary electrifiers 119a and 119b accompanying the switching of the scanning direction of the wafer is done by the preliminary electrification controller 120 through the command of the inspection apparatus controller 117. In the case where a wafer to be inspected is likely to maintain an electrified state, it is possible to operate the preliminary electrifiers 119 intermittently during the inspection. The region, electrified with the preliminary electrifiers 119 is sufficiently larger than the region of inspection and thus it is possible to electrify a large area of region by scanning the wafer only once in the X direction. Therefore, it is possible to stop the operation of the preliminary electrifiers 119 in the successive scanning. The inspection apparatus controller 117: decides the timing of the turn-on and turn-off of the preliminary electrifiers 119 in accordance with the setting by a user; and instructs the preliminary electrification controller 120. The details of one of the preliminary electrifiers 119 are shown in FIG. 3. FIG. 3a shows a cross section of a preliminary electrifier. This is the case where the irradiation beam used for preliminary electrification is an electron beam. As the electron beam source, an areal electron source 301 represented by an electron source using carbon nanotubes (CNT) is used. The electron source 301 is disposed in the vacuum chamber 201 through an electrically insulated table 302 and a cable to apply the acceleration voltage Va of the electron beam is connected from the outside of the vacuum through an introduction terminal 303. A grid-shaped extraction electrode 304 is disposed while being insulated from the vacuum chamber 201 and a cable to apply the extraction voltage Ve is led to the outside of the vacuum through another introduction terminal 303. Although the electron source 301 and the extraction electrode 304 are attached to the vacuum chamber 201 independently from each other in the figure, they may be used as a structure formed by integrating them. An electrification control electrode 305 to control the electrification of a wafer 104 is disposed so as to be as close to the wafer 104 as possible while being insulated from the vacuum chamber. In the case of such an areal electron source as a CNT electron source, the parallelism of a beam is good and hence, when a grid-shaped electrode is used, a portion which is shaded by the grid and shields the permeation of an electron beam reaches the wafer as it is, and a grid-shaped unevenness of irradiation beam strength is caused. The unevenness of the distribution in the X direction is equalized since the wafer moves in the X direction during inspection, but the unevenness in the direction (the Y direction) perpendicular to the moving direction is not equalized. In mirror electron imaging wherein uniformity of electrification potential is required, such unevenness of the irradiation electron beam strength induces abnormality of the contrast of an image and causes misinformation to be obtained. FIG. 3b shows a structure of the electrification control electrode 305. The electrification control electrode 305 has a structure formed by providing an electrically-conductive plate with an opening having the width Wc in the direction (the X direction) wherein the stage moves during inspection and the length Lc in the direction (the Y direction) perpendicular to the direction wherein the stage moves during inspection (the X direction). The length Lc is designed so as to be longer than the width Wc. A cable to apply the voltage Vc to the electrification control electrode 305 is led to the outside of the vacuum through another introduction terminal 303. The voltages applied to the electrodes and the like are supplied from the preliminary electrification controller 120 in accordance with the setting of a user. Further in the present embodiment, a beam forming slit 306 is provided between the extraction electrode 304 and the electrification control electrode 305. FIG. 3c shows a structure of the beam forming slit 306. The beam forming slit 306 has a structure formed by providing an electrically-conductive plate with an opening having the width Ws and the length Ls, and maintains the relationship with the width Wc and the length Lc of the opening of the electrification control electrode 305 so as to satisfy the expressions Ws<Wc and Ls<Lc. The purpose of the installation of the beam forming slit 306 is to prevent electrons coming from the electron source 301 from colliding with the electrification control electrode 305. By the installation of the beam forming slit 306, it is possible to prevent the danger that secondary electrons generated by the exposure of the electrification control electrode 305 to an electron beam disturb the electrification potential formed on the wafer 104. Further, if a heating mechanism to prevent contamination caused by the electron beam irradiation to the beam forming slit 306 is provided, it is possible to prevent the abnormal deflection of an electron beam caused by the electrification of an adsorptive material and maintain uniform irradiation to a wafer. By the means of the present embodiment, it is possible to always apply uniform electron beam irradiation to a wafer in preliminary electrification operation, and hence it is possible to form a uniform electrification potential and increase the accuracy of wafer inspection. In the present embodiment, an example of an extraction electrode having a structure other than that of a usually used grid electrode in a preliminary electrifier 119 in the apparatus configuration shown in FIG. 1 is explained. FIG. 4 shows a structure of the extraction electrode. The extraction electrode has a lattice structure formed by aligning extra-fine conductive wires 402 in parallel on an opening 401 through which an electron beam passes. The direction of the extension of the wires 402 is the direction (the Y direction) perpendicular to the direction of the movement of a wafer stage during inspection. By mounting the electrode of the present embodiment onto a preliminary electrifier shown in FIG. 1 or FIG. 3a, it is possible to apply electron beam irradiation which is uniform in the direction perpendicular to the direction of the wafer stage movement during inspection to a wafer from the stage of the extraction of the electron beam. As a result, it is possible to form an electrification potential having enhanced uniformity and further increase the accuracy of wafer inspection. There are some cases where a large amount of electric charge must be supplied to a wafer until a desired electrification potential is obtained in accordance with the type of a wafer to be inspected. Since the speed of the wafer stage cannot be changed in the preliminary electrification during inspection, in order to supply a larger amount of electric charge, it is necessary to take a means of either increasing the electric current from an electron source or expanding the irradiation area to the direction of the wafer stage movement. When electric current from an electron source is increased, disadvantages including the shortening of the service life of the electron source and the like tend to arise. To cope with the problem, in the present embodiment, adopted is a structure wherein, in such a preliminary electrifier as shown in FIG. 1 or FIG. 3a, an electrification control electrode 501 and a beam forming slit 502, those being provided with plural openings, are used and thus the irradiation area of a beam is expanded. FIG. 5 shows the structures of the electrification control electrode 501 and the beam forming slit 502. The intervals P (the distance P between the starting point A of an opening and the starting point B of the adjacent opening in the direction of the stage movement (the X direction)) of the openings of the electrification control electrode 501 are set to be equal to the intervals P of the openings of the beam forming slit 502, and thereby an electron beam can be adjusted by using the beam forming slit 502 so as to pass through only the inside of the openings of the electrification control electrode 501. By applying the present embodiment to the apparatus shown in FIG. 1, it is possible to increase the amount of an electron beam supplied to a wafer while maintaining the uniformity of the electrification potential without imposing burdens on an electron source. In the present embodiment, described is an example wherein the amount of electrons with which a wafer is irradiated is reduced gradually toward the outside in a preliminary electrifier 119 of an apparatus shown in FIG. 1 or FIG. 3a. Conventionally, a preliminary electrifier has been disposed in the vicinity of an electron optical column for imaging in a mirror electron imaging type inspection apparatus. A wafer to be inspected moves little by little in the Y direction while reciprocally moving in the X direction and the entire surface of the wafer is inspected. The region to which preliminary electrification is applied must be sufficiently larger than the field of view of inspection in order to prevent the boundary of the preliminary irradiation region from affecting the potential distribution in the field of view of mirror electron imaging, namely in order to form a uniform electrification potential in the inspection region. For example, whereas the size of the field of view in mirror electron imaging during inspection is 100 to 200 microns, the size of the region of the preliminary irradiation is about 10 mm, and thus the sizes are largely different from each other. As shown in FIG. 16, when a wafer moves continuously in the X direction, the region to which preliminary electrification is applied extends in the form of a strip on the surface of the wafer. A part to which preliminary irradiation is already applied is electrified to a prescribed potential and a part to which preliminary irradiation is not applied has a potential different from the part to which preliminary irradiation is already applied. In such a state, since the width of the strip is larger than the field of view of inspection, the boundary of the electrification region reaches an uninspected region which has yet to be inspected. Since a mirror electron imaging type inspection apparatus forms a potential distribution on a surface as an image, when the boundary of a preliminary electrification region having potential gradient is observed as it is with mirror electrons, the boundary forms a line and undesirably appears as an image. The linear contrast appearing in the mirror electron image strengthens as the steepness of the change of the electric charge distribution in the boundary region increases. For example, in the case of a wafer which is coated with a material wherein the time of diffusion of electrified charge is long, the steep change of a potential at the boundary of the region electrified by preliminary electrification is undesirably maintained. Although preliminary electrification is newly applied immediately before the pattern overlapping with the boundary region is inspected, it takes time until already generated difference in the amount of electrified charge is completely eliminated and it happens that an image is obtained while the line remains. In such a case, even when preliminary electrification is applied immediately before inspection, the boundary remaining as a result of the preliminary electrification applied in the previous inspection appears in a mirror electron image as contrast abnormality when the subsequent inspection image is obtained. A problem has been that, by such abnormal contrast in a mirror electron image caused by the unevenness of electrification potential, a portion which is not a defect is undesirably counted as a defect and correct inspection cannot be carried out in actual inspection. Then in the present embodiment, in a preliminary electrifier 119 of an apparatus shown in FIG. 1 or FIG. 3a, the opening of a beam forming slit 601 is provided with opening width changed portions 602 wherein the opening width narrows toward both the ends as shown in FIG. 6a. By forming the opening into such a shape, it is possible to: decrease the amount of electrons with which a wafer is irradiated gradually toward the outside; decrease the gradient of the electrification potential change at the electrification potential boundary as shown in FIG. 6c; and prevent contrast caused by a steep potential gradient from appearing. Further, a beam forming slit 603 shown in FIG. 6b is provided with aligned plural openings having a shape shown in FIG. 6a, and by using such a slit, it is possible to increase the amount of electric charge supplied to a wafer without generating a steep potential portion at the boundary. Here, in any of the cases, the electrification control electrode may be provided with a rectangular opening or rectangular openings as shown in FIG. 5. In Embodiment 4, an opening of a beam forming slit is formed so as to narrow toward both the ends thereof. However, when plural openings are fabricated as shown in FIG. 6b, the fabrication process becomes complicate. To cope with that, in the present embodiment, the lengths of plural openings are changed little by little as shown in FIG. 7 in a preliminary electrifier 119 of the apparatus shown in FIG. 1 or FIG. 3a. By the means of the present embodiment, it is possible to: increase the amount of electric charge supplied to a wafer without generating a steep potential part at a boundary; and realize preliminary irradiation that does not cause steep potential change through easier fabrication process. In the examples explained above, an extraction electrode to extract preliminary irradiation electrons is provided. In the present embodiment, such a slit 701 as shown in FIG. 7 is disposed in the vicinity of an electron source instead of an extraction electrode, and thereby a beam forming slit and an extraction electrode are combined as shown in FIG. 8 in a preliminary electrifier 119 of the apparatus shown in FIG. 1. By adopting such arrangement, it is possible: not only to simplify the structure; but also to eliminate the unevenness of the irradiation electron distribution caused by the grid of an extraction electrode from the bottom up and further equalize the amount of an irradiation electron beam. In the present embodiment, a blanker 901 is disposed between a slit combined with an extraction electrode as shown in the invention of Embodiment 6 and an electrification control electrode as shown in FIG. 9 in a preliminary electrifier 119 of the apparatus shown in FIG. 1. The preliminary irradiation is applied during inspection. However, in the case of a wafer having sufficiently long electrification retention time, it is not necessary to apply the preliminary irradiation continuously during inspection and the preliminary irradiation may be stopped during pre-measured retention time. That is, the preliminary irradiation is applied intermittently at intervals of time during inspection. On this occasion, if the generation itself of an electron beam from an electron source is stopped, it sometimes happens that the state of a power source and the state of the electron source slightly change when a subsequent electron beam is generated and the same irradiation strength as generated before the stop is not reproduced. Further, repeated turn-on and turn-off of the electron source may cause the service life of the electron source itself to shorten. To cope with that, in the present embodiment, as a means of stopping irradiating a wafer with an electron beam without stopping generating electrons from an electron source, a blanker 901 is disposed. When the irradiation of an electron beam to a wafer is stopped, a predetermined voltage is applied to the blanker 901, the route of the electron beam is largely deflected, and thereby the electron beam is shielded with the sidewall of the blanker 901. The timing of the irradiation is controlled with the preliminary electrification controller 120. By the means of the present embodiment, it is possible to intermittently irradiate a wafer always with a constant amount of electron beam without deteriorating the service life of the electron source. In the present embodiment, as shown in FIG. 17, a movable plate 1072 having such plural prescribed openings 1701 as shown in aforementioned examples beforehand is used as a beam forming slit in a preliminary electrifier 119 of an apparatus shown in FIG. 1 or FIG. 3a. When only wafers 1703 which have an identical chip size and do not have large difference in time required for preliminary electrification are inspected, it is not necessary to frequently change the size of the openings. However, when a chip having a different size or a wafer which requires longer time for preliminary electrification is successively inspected, openings suitable for the new wafer are selected, the openings are moved to intended places, and the inspection is carried out. It is possible to adopt a method of moving a wafer little by little in the Y direction while reciprocally moving the wafer in the X direction so as to be able to inspect the entire wafer during inspection. With regard to the movement in the Y direction however, it is also possible to move the movable plate 1702 with a drive unit in synchronization with the movement of a wafer 1703 under inspection. This kind of control is carried out with the preliminary electrification controller 120. By adopting such a configuration, it is possible to: change the size of an irradiation region merely by selecting openings prepared in a movable plate; and thus increase the inspection speed. The above examples have been based on the premise that an electron beam is used as the irradiation beam for preliminary electrification. It is possible to control the electrification of a wafer also by the irradiation of ultraviolet rays in the same way as the electron beam irradiation. FIG. 10 shows the case where ultraviolet rays are used as an irradiation beam source other than an electron source in a preliminary electrifier 119 of such an apparatus as shown in FIG. 1. An ultraviolet light source 1001 generates ultraviolet rays having a sufficient energy to excite photoelectrons of a wafer. The ultraviolet light source 1001 is controlled with a controller 1002, and the controller carries out the setting of the ultraviolet light strength and the irradiation time, turn-on and turn-off of the ultraviolet light, and others in accordance with instructions of the preliminary electrification controller 120. Such ultraviolet rays are generally vacuum ultraviolet rays that do not permeate the air and hence the light source is placed in a vacuum chamber. An outer casing 1003 to shield the part not requiring ultraviolet light from irradiation is disposed and an electrification control slit 1004 is disposed at the tip thereof protruding toward the side of a wafer. By the means of the present embodiment, it is possible to electrify a wafer with ultraviolet rays which are more stable than an electron beam. In the case of Embodiment 9, ultraviolet rays diffuse from an ultraviolet light source, the amount of the ultraviolet rays which can actually contribute to electrification is small, and there is a possibility that electric charge necessary for the desired electrification of some wafers cannot be generated. To cope with the problem, in the present embodiment, a reflex mirror 1101 is disposed on an ultraviolet light source so that the diffused ultraviolet rays may be condensed and may reach a wafer as shown in FIG. 11. By applying the means of the present embodiment to such an apparatus shown in FIG. 1, it is possible to: efficiently irradiate a wafer with ultraviolet light; and electrify a wafer at a shorter period of time. In the above examples, the cases where wafers are electrified simultaneously during inspection to detect electrical defects have mainly been explained. In those cases however, since the beam irradiation time is restricted by the traveling speed of a wafer, the setting of a long beam irradiation time is also restricted. Further, in some semiconductor manufacturing processes, an electrification distribution having a large potential is caused on the whole wafer. Such a large electrification potential cannot be eliminated with the preliminary irradiation applied simultaneously with the inspection which allows only a limited irradiation time and there is a possibility that large contrast abnormality arises in a mirror electron image. To cope with the problem, in the present embodiment, when a wafer is transferred from a preliminary chamber 202 to a vacuum chamber 201, the already existing electrification of the wafer is eliminated or reduced to a potential level controllable by preliminary electrification applied at inspection with a preliminary static eliminator 1201 as shown in FIG. 12. A desirable place where the preliminary static eliminator 1201 is installed is on the side of the vacuum chamber 201 at the opening between the preliminary chamber 202 and the vacuum chamber 201. By adopting such arrangement, it is possible to eliminate electrification by preliminary electrification during the transportation of a wafer and hence the inspection time is not substantially disturbed. The basic configuration of a preliminary static eliminator 1201 is shown in FIG. 13. FIG. 13 is a view taken by viewing a wafer charging port 1304 on the side of a preliminary chamber from the main chamber of an apparatus and a wafer 104 is placed on a transfer table 1303 and moves from the rear side of the figure to the front side thereof. An electron source 1301 which is longer than the diameter of the wafer is used in order to uniformly irradiate the entire surface of the wafer with an electron beam. The electron source 1301 comprises carbon nanotubes built on a rectangular substrate for example, and, though it is not shown in the figure, an extraction electrode of a grid-shape is disposed in the vicinity in order to extract electrons from the carbon nanotubes and an electron beam of a rectangular cross section can be generated. An electrification control electrode 1302 is disposed right above the wafer. The electrification control electrode comprises a rectangular opening and a mesh and forms a potential of nearly zero volt immediately above the wafer. Unlike the preliminary electrifiers of the previous examples, it is not necessary to form a uniform electrification distribution immediately before inspection, it is only necessary to control the potential of a wafer to the extent of allowing being recreated by preliminary electrification during inspection, and hence a mesh electrode is used for the electrification control electrode. Details such as structures to support members, cables to supply power source to electrodes, and others are omitted here. The voltages applied to the electron source and the electrodes are controlled and supplied with a preliminary static elimination controller 1305 which is controlled with an inspection apparatus controller 117. Besides such a rectangular electron source of a large size as described in the above example, it is possible to configure an electron source of a preliminary static eliminator 1201 by using such an electron source as used in a preliminary electrifier 119. As shown in FIG. 14, plural electron sources 1401 each of which is provided with an extraction electrode are disposed in the form of staggered rows comprising at least two rows so that the irradiation regions may overlap with each other. When a sufficient amount of electron beam irradiation is required, the number of the parallel electron source rows is increased appropriately. In this case, besides the electron sources 1401, vacuum ultraviolet light sources aligned in plurality may also be used. FIG. 15 shows the major parts of a flow of wafer inspection to which the means of the present embodiment is applied. Though not shown in the figure, it is assumed that various conditions of each operation are transferred from an inspection apparatus controller to each controller beforehand or properly. When a wafer is introduced from a wafer transfer robot in the state where a preliminary chamber is open to the air, a wafer charging port of the preliminary chamber is sealed and the air is evacuated. At the same time, the operation of preliminary static elimination is started to prepare for the static elimination during the transportation of the wafer. When the degree of vacuum in the preliminary chamber comes to a sufficiently low level, a gate valve between the preliminary chamber and an inspection apparatus main chamber opens and the transportation of the wafer is started. The traveling speed during the wafer transportation is properly adjusted so as to secure sufficient static elimination. When the wafer is completely transferred into the apparatus main chamber, the beam of the preliminary static eliminator is stopped or the wafer is shielded from the beam, the beam irradiation is finished, and simultaneously the gate valve is closed. The wafer waits for the commencement of inspection at a home position (HP) and successively moves to a position where next inspection starts. Thereafter, the operation of the preliminary electrifier starts, and inspection is commenced on condition that a sufficient and stable irradiation beam strength and voltage conditions conforming to the recipe are obtained. When the inspection of the wafer under preliminary electrification is finished, the preliminary electrifier stops the beam or shields the wafer from the beam, and the wafer is returned to the home position (HP) again. When the wafer is not reinspected, the preliminary static eliminator restarts the operation toward the conditions allowing the beam irradiation of a wafer. At this moment, the electrode conditions are set so that the potential of the wafer may be zero volt. When the degree of vacuum in the preliminary chamber is sufficiently low, the gate valve between the apparatus main chamber and the preliminary chamber opens and the static elimination operation and transportation of a wafer are carried out. After the wafer is completely transferred into the preliminary chamber, the beam of the preliminary static eliminator stops or the wafer is shielded. The gate valve between the apparatus main chamber and the preliminary chamber is closed, the preliminary chamber is open and released to the air, and thereafter the wafer is extracted from the preliminary chamber with the transfer robot and returns to the wafer charging port. At this stage, the potential of the electrification of the wafer is uniform and nearly zero volt, and the next process is not influenced at all. By the means of the present embodiment, even when unintended electrification caused by various processes is formed on a wafer beforehand, it is possible not only to carry out stable and accurate inspection but also to remove the influence of the electrification on the next processes. The examples according to the present invention have been described above. The combinations of those examples are also included in the present invention. The present invention includes a defect inspection method which uses a defect inspection apparatus: to obtain an image of a circuit pattern formed on a specimen by areally irradiating a first region of a specimen introduced from an entrance with a first electron beam, reflecting the first electron beam immediately before falling on the specimen, and focusing the reflected electrons into an image; and to detect defects existing in the circuit pattern on the basis of the obtained circuit pattern image, wherein the method comprises the processes of: equalizing the electrification distribution in the first region by irradiating a second region including the first region on the specimen surface with ultraviolet rays or a second electron beam before the first region is inspected; and irradiating the specimen surface with a third electron beam or ultraviolet rays in the vicinity of the entrance. The present invention further includes a defect inspection method wherein, in the aforementioned defect inspection method, the region irradiated with the third electron beam or the ultraviolet rays is longer than the diameter of the specimen. The present invention further includes a defect inspection method wherein, in either of the aforementioned defect inspection methods, plural electron sources or plural ultraviolet light sources are used for the irradiation with the third electron beam. The present invention further includes an inspection apparatus: to obtain an image of a circuit pattern formed on a specimen by areally irradiating a first region of the specimen with a first electron beam nearly in parallel, reflecting the first electron beam immediately before falling on the specimen, and focusing the reflected electrons into an image; and to detect defects existing in the circuit pattern on the basis of the obtained circuit pattern image, wherein: the inspection apparatus is provided with the means of irradiating a second region including the first region on the specimen with ultraviolet rays or a second electron beam before the circuit pattern image is obtained; and the deviation of each electrification potential in the electrification potential distribution in the first region from the average value of the electrification potentials on the entire surface of the specimen is 1 V or less. The present invention further includes an inspection apparatus which is provided with: an electron optical system to irradiate a first region of a specimen introduced from an entrance with an electron beam; a specimen stage to retain the specimen; a means of applying such a voltage that the electron beam with which the specimen is irradiated does not enter the specimen but is reflected to the specimen stage or the specimen; a means of detecting the electrons reflected from the side of the specimen by the application of the voltage; a means of forming an inspection image and detecting a defect of the specimen on the basis of the detection signals of the detecting means; a means of irradiating the first region with ultraviolet rays or a second electron beam before the inspection image is formed; and a means of irradiating the specimen with a third electron beam or vacuum ultraviolet rays in the vicinity of the entrance. The present invention further includes an inspection apparatus wherein, in the aforementioned inspection apparatus, the means of the irradiation with the third electron beam is provided with plural electron sources aligned in parallel. |
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claims | 1. An ultra-cold-matter (UCM) system comprising:a hermetically-sealed ultra-high-vacuum (UHV) enclosure;a source cell having a source-cell enclosure, the source cell being nested within the UHV enclosure, the source cell being physically attached to and separated from the UHV enclosure so as to be thermally isolated from the UHV enclosure;a source material in a non-vapor phase and disposed within the source cell, said source material being characterized in that vapor-phase source atoms can be released from the source material;a cold-atom trap for cooling at least some of the vapor-phase source atoms to an ultra-cold temperature; andan atom getter for maintaining an ultra-high vacuum within the UHV enclosure at least in part by causing at least some of the vapor-phase source atoms to be sorbed to or into a getter material, the atom getter being disposed within the UHV enclosure and outside the source cell. 2. The UCM system of claim 1 further comprising a laser external to the UHV enclosure that serves as a source of a laser light, the UHV enclosure including a first transparent material that is transparent to the laser beam, the source-cell enclosure including second transparent material that is transparent to the laser beam and that is more resistant to corrosion by the vapor-phase source atoms than is the first transparent material. 3. The UCM system of claim 2 wherein the vapor-phase source atoms are atoms of strontium, the first transparent material is glass and the second transparent material is sapphire. 4. The UCM system of claim 1 wherein the getter material is disposed within the source cell and serves to regulate the amount of vapor-phase source atoms in the source cell by sorbing and releasing source atoms as a function of the partial pressure of the source atoms in the source cell. 5. The UCM system of claim 4 wherein the getter material includes at least one of gold, carbon, and antimony. 6. The UCM system of claim 1 further comprising:a first light-reflecting element set of at least a first reflecting element within the source-cell enclosure arranged to support a two-dimensional atom trap;a second light-reflecting element of at least a second reflecting element located within the enclosure and external to the source-cell enclosure and arranged to support a three-dimensional atom trap. 7. The UCM system of claim 1 wherein the atom getter includes an ion pump that ionizes the source atoms to yield source ions, the ion pump accelerating the source ions so that they contact that getter material. 8. The UCM system of claim 1 further comprising a hot mirror disposed between the source-cell enclosure and the UHV enclosure. 9. The UCM system of claim 8 wherein the hot mirror is formed as a coating on a surface of the source-cell enclosure facing the UHV enclosure. 10. The ultra-cold-matter system of claim 1 further comprising:a heater element disposed within the source-cell interior; andelectrical feed-throughs through the source cell and the UHV enclosure, the electrical feed-throughs providing electrical paths to and from the heater element. 11. The UCM system of claim 1 wherein the source material is an alkaline earth metal and the source cell is predominantly of sapphire or high-alumina-content silicate glass. 12. A process comprising:generating a vapor phase of source particles at least in part by heating a non-vapor-phase material, the generating occurring within a source cell nested within a vacuum enclosure;using a vacuum established between the vacuum enclosure and the source cell, thermally isolating an ultra-cold region within the vacuum enclosure and external to the source cell from heat generated in the source cell;using a laser, pre-cooling source particles within the source cell;transferring pre-cooled source particles from the source cell to the ultra-cold region; andusing a laser, trapping the source particles in the ultra-cold region. 13. The process of claim 12 further comprising regulating a pressure of the source particles in the source cell using getter material in the source cell that sorbs and releases at least some of the source particles. 14. The process of claim 12 further wherein the thermally isolating includes reflecting infrared light exiting the source cell using a hot mirror. 15. The process of claim 12 further wherein the source cell is nested within the vacuum enclosure using standoffs to spatially and thermally isolate the source cell from the vacuum enclosure. 16. The process of claim 12 further wherein the transferring includes gettering off-axis particles between the source cell and the ultra-cold region. |
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abstract | A safety valve drive system is operated such that a safety valve of a nuclear power plant is opened by supplying a driving gas by using a pilot valve at an occurrence of an accident or a transient state to thereby protect a reactor against pressure application. The safety valve drive system is provided with a safety valve drive unit, as a function of actuating the safety valve, and cables. The safety valve drive unit actuates in a manner that the safety valve is opened in response to respective auto-depressurization system actuating signals for two or more segments among respective auto-depressurization system actuating signals for four segments, and is closed if an auto-depressurization system actuating signal for one or less segment among the auto-depressurization system actuating signals for the four segments is received. The cables are connected to the safety valve drive unit and used to transfer the auto-depressurization system actuating signals for the four segments. |
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claims | 1. A method for determining an operating margin for a nuclear fuel core of a particular Boiling Water Reactor (BWR), wherein fuel design or core configuration are contingent upon an operating margin for the reactor, said operating margin being determined by a process for evaluating an operating limit minimum critical power ratio (OLMCPR), comprising: a) determining a generic transient change-in-critical-power-ratio (xcex94CPR/ICPR) distribution of values, wherein said generic transient xcex94CPR/ICPR distribution is based at least upon a pre-computed nominal xcex94CPR/ICPR bias value and an xcex94CPR/ICPR uncertainty; b) determining initial critical power ratios (ICPRs) for each fuel rod in the core corresponding to a particular BWR plant; c) determining a plurality of transient Minimum CPRs (MCPRs) for all fuel rods in the fuel core by applying the generic transient xcex94CPR/ICPR distribution to each ICPR determined in (b); and d) effecting said Boiling Water Reactor operation by applying a selected OLMCPR as an operational control parameter, said OLMCPR selected for the particular BWR by computing a mean NRSBT (number of fuel rods subject to boiling transition) corresponding to a histogram of said plurality of transient Minimum CPRS and selecting an OLMCPR value as a transient Minimum CPR which results in a corresponding mean NRSBT value that is less than a predetermined cutoff value. 2. The method of claim 1 wherein the pre-computed nominal xcex94CPR/ICPR bias value and uncertainty is determined by conventional statistical analysis of one or more computer simulations of at least one type of reactor transient operational occurrence for a least one class of BWR plant type and for at least one fuel type. claim 1 3. The method of claim 1 wherein said ICPRs are determined for a particular transient event during a particular fuel cycle. claim 1 4. A method for determining an operating margin for a nuclear fuel core of a particular Boiling Water Reactor (BWR), wherein fuel design or core configuration are contingent upon an operating margin for the reactor, said operating margin being determined by a process for evaluating an operating limit minimum critical power ratio (OLMCPR), comprising: determining initial critical power ratios (ICPRs) for each fuel rod in a fuel core of the particular BWR; using a predetermined generic transient change-in-critical-power-ratio (xcex94CPR/ICPR) distribution of values to compute transient Minimum CPRs (MCPRs) for all fuel rods in the fuel core based upon said ICPRs; computing a mean NRSBT (number of fuel rods subject to boiling transition) corresponding to a histogram of said transient Minimum CPRs; selecting an operating limit minimum critical power ratio OLMCPR) for the BWR from a histogram of fuel rod transient Minimum CPUs for a particular set of initial conditions that result in a mean NRSBT value that is less than a predetermined cutoff value. 5. The method of claim 4 wherein e predetermined generic transient xc3x97CPR/ICPR distribution of values is determined by statistical analysis of one or more computer simulations of at least one type of reactor transient operational occurrence for a least one class of BWR plant type and for at least one fuel type. claim 4 |
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summary | ||
claims | 1. A heat exchanger module with longitudinal axis (X) comprising a stack of plates defining at least two fluid circuits, at least some the plates each comprising fluid circulation channels each delimited at least in part by a groove, the channels of at least one of the two circuits, termed the first circuit, including:a zone (Z1), termed the feeding zone, of feeding the fluid from the exterior of the stack, in which the channels are parallel to one another and extend along a secant axis (X′) intersecting the longitudinal axis (X) and in which two adjacent channels communicate with one another via at least one notch formed in a rib separating their respective grooves;a zone (Z3) termed the bifurcation zone in which each channel is divided into at least two straight channels parallel to one another and parallel to the longitudinal axis (X), being separated from one another by a rib;a zone (Z2) termed the connection zone between the feeding zone and the bifurcation zone, in which each channel has a straight profile that extends along the secant axis (X′) and a curved profile continuous with the straight profile in order to connect the channel with a straight channel of the bifurcation zone;a zone (Z4) of continuous exchange with the bifurcation zone in which the parallel straight channels separated from one another by ribs extend parallel to the longitudinal axis (X);wherein the channels of each plate of the first circuit communicate with those of the other plates of the first circuit in their respective feed zone (Z1), via openings, made in each channel of the feeding zone, passing through the stack but not communicating with the channels of the second circuit,the notches and the openings forming a jet-break grille for rebalancing the flows of the fluid between the channels of the first circuit when said fluid is circulating in said channels. 2. The heat exchanger module as claimed in claim 1, wherein the curved profile of each channel of the first circuit comprises two curves to connect the straight profile of the connection zone to the straight channel of the bifurcation zone. 3. The heat exchanger module as claimed in claim 1, wherein each straight channel is divided into four channels in the bifurcation zone (Z3). 4. The heat exchanger module as claimed in claim 1, wherein the angle between the secant axis (X′) and the longitudinal axis (X) of the module is between 0 and 45° inclusive. 5. The exchanger module as claimed in claim 1, wherein a plate of the first circuit is inserted between two plates of the other of the two circuits, termed the second circuit, at least in the central part of the stack. 6. The exchanger module as claimed in claim 1, wherein the channels of the first circuit have an oval, circular, rectangular or square section. 7. A method of producing a heat exchanger module as claimed in claim 1, comprising the following steps:machining grooves in first metal plates in order to constitute the channels of the first circuit configured with the feed, connection, bifurcation and exchange zones;machining grooves in second metal plates in order to constitute the channels of other of the two circuits, termed the second circuit;stacking in an alternating manner the first plates and the second plates so as to have the through-openings that enable communication between channels of the plates of the first circuit but not with those of the plates of the second circuit;assembling the first and second metal plates to one another, either by hot isostatic compression (HIC), or by a process termed a hot uniaxial diffusion welding process, so as to obtain welding by diffusion between them, or by brazing. 8. The use of a heat exchanger comprising a plurality of heat exchanger modules as claimed in claim 1, wherein the fluid of the first circuit, by way of primary fluid is a liquid metal and the fluid of a second circuit, by way of a secondary fluid, being a gas or a gas mixture. 9. The use as claimed in claim 8, wherein the fluid of the second circuit comprises mainly nitrogen and the fluid of the first circuit being liquid sodium. 10. The use as claimed in claim 8, wherein the fluid of the first or second circuit comes from a nuclear reactor. 11. A nuclear installation comprising a fast neutral nuclear reactor cooled with liquid metal, notably a liquid sodium cooled fast reactor (SFR) and a heat exchanger comprising a plurality of exchanger modules as claimed in claim 1. |
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claims | 1. A computer implemented method for providing general data protection regulation (GDPR) compliant hashing in blockchain ledgers, the method comprising:receiving a first message from a user at a blockchain gateway device, wherein the first message comprises personal identification information (PII);performing, at the blockchain gateway device, a first hashing function on the first message to obtain a hash value of the first message;storing the hash value of the first message in a blockchain ledger;storing the first hashing function in an off-chain database;storing the first message in the off-chain databasesreceiving, at the blockchain gateway device, a request to delete the first message;arbitrarily selecting a second message that is different than the first message;calculating a second hashing function using the second message, wherein the second hashing function results in the same hash value;replacing the first message in the off-chain database with the second message;replacing the first hashing function in the off-chain database with the second hashing function; andstoring a GDPR compliant hash value of the message in the blockchain ledger. 2. The method of claim 1, wherein the personal identification information (PII) of the user is selected from social security numbers, mailing addresses, email addresses, phone numbers, IP addresses, login IDs, social media posts, digital images, geolocation data, biometric data, and behavioral data. 3. The method of claim 1, further comprising:requesting, by a data controller, a GDPR proof of compliance to the blockchain gateway device; andproviding the GDPR compliant message to the data controller in response to the request. 4. The method of claim 1, wherein the GDPR compliant message comprises a timestamp. 5. The method of claim 1, wherein the first hashing function and the second hashing function comprise a selected prime number. 6. The method of claim 1, wherein performing, at the blockchain gateway device, a first hashing function on the first message to obtain a hash value of the first message further comprises:selecting a prime number p and a primitive root g mod p;letting the first message be represented by a first residue m mod p;selecting a first hashing function as a second residue s mod p; andwherein storing the first message in the off-chain database comprises storing the first residue m mod p, storing the first hashing function in the off-chain database comprises storing the second residue s mod p, and storing the hash value of the first message in the blockchain ledger comprises storing the third residue v≡ms mod p. 7. The method of claim 6, further comprising:wherein receiving the request to delete the first message comprises a request to delete the first residue of m;wherein arbitrarily selecting a second message that is different than the first message further comprises selecting mnew mod p;wherein calculating a second hashing function using the second message, wherein the second hashing function results in the same hash value further comprises calculating snew≡vmnew−1 mod p; andwherein replacing the first message in the off-chain database with the second message further comprises replacing m in the off-chain database with mnew;wherein replacing the first hashing function in the off-chain database with the second hashing function further comprises replacing s with snew. 8. The method of claim 6, wherein the prime number p and the primitive root g mod p can both be made public. 9. A blockchain gateway device for providing general data protection regulation (GDPR) compliant hashing in blockchain ledgers, the system comprising:a processor; andone or more memory devices storing computer-executable instructions that, when executed with the processor, cause the system to at least:receive a first message from a user, wherein the first message comprises personal identification information (PII);perform a first hashing function on the first message to obtain a hash value of the first message;store the hash value of the first message in a blockchain ledger;store the first hashing function in an off-chain database;store the first message in the off-chain database;receive a request to delete the first message;arbitrarily select a second message that is different than the first message;calculate a second hashing function using the second message, wherein the second hashing function results in the same hash value;replace the first message in the off-chain database with the second message;replace the first hashing function in the off-chain database with the second hashing function; andstore a GDPR compliant hash value of the message in the blockchain ledger. 10. The system of claim 9, wherein the personal identification information (PII) of the user is selected from social security numbers, mailing addresses, email addresses, phone numbers, IP addresses, login IDs, social media posts, digital images, geolocation data, biometric data, and behavioral data. 11. The system of claim 9, wherein the one or more memory devices storing computer-executable instructions that, when executed with the processor, cause the system to further:receive a request from a data controller for GDPR proof of compliance; andprovide the GDPR compliant message to the data controller in response to the request. 12. The system of claim 9, wherein the GDPR compliant message comprises a timestamp. 13. The system of claim 9, wherein the first hashing function and the second hashing function comprise a selected prime number. 14. The system of claim 9, wherein the one or more memory devices storing computer-executable instructions that, when executed with the processor, cause the system to further perform a first hashing function on the first message to obtain a hash value of the first message further comprises computer-executable instructions to:select a prime number p and a primitive root g mod p;let the first message be represented by a first residue m mod p;select the first hashing function as a second residue s mod p; andstore the first residue m mod p as the first message;store second residue s mod p as the first hashing function; andstore third residue v≡ms mod p as the hash value. 15. The system of claim 14, wherein the one or more memory devices storing computer-executable instructions that, when executed with the processor, further comprises computer-executable instructions to:receive a request to delete the first residue of m;select a second message that is different than the first message as mnew mod p;calculate a second hashing function using the second message, wherein the second hashing function results in the same hash value, as snew≡vmnew−1 mod p; andreplace the first message in the off-chain database with the second message by replacing m in the off-chain database with mnew;replace the first hashing function in the off-chain database with the second hashing function by replacing s with snew. 16. The system of claim 14, wherein the prime number p and the primitive root g mod p can both be made public. 17. One or more non-transitory computer-readable media having computer-executable instructions for performing a method of running a software program on a computing device for providing general data protection regulation (GDPR) compliant hashing in blockchain ledgers, the method comprising, the computing device operating under an operating system, the method including issuing instructions from the software program comprising:receiving a first message from a user at a blockchain gateway device, wherein the first message comprises personal identification information (PII);performing, at the blockchain gateway device, a first hashing function on the first message to obtain a hash value of the first message;storing the hash value of the first message in a blockchain ledger;storing the first hashing function in an off-chain database;storing the first message in the off-chain database receiving, at the blockchain gateway device, a request to delete the first message;arbitrarily selecting a second message that is different than the first message;calculating a second hashing function using the second message, wherein the second hashing function results in the same hash value;replacing the first message in the off-chain database with the second message;replacing the first hashing function in the off-chain database with the second hashing function; andstoring a GDPR compliant hash value of the message in the blockchain ledger. |
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claims | 1. An automated method for repairing spacing of vertical conductive features in the fabrication of semiconductor integrated circuits, the method comprising:determining a minimum spacing;using a database, identifying vertical conductive feature patterns having a spacing less than the minimum spacing;for each identified vertical conductive feature pattern having a spacing less than the minimum spacing, determining a first direction to expand and a second direction to shrink to define a revised vertical conductive feature pattern;checking the revised vertical conductive feature pattern against design rules to see if the design rules are violated; andfor each identified vertical conductive feature pattern having a spacing less than the minimum spacing, replacing the identified vertical conductive feature pattern with the revised vertical conductive feature pattern when the design rules are not violated, wherein the revised vertical conductive feature pattern is expanded in the first direction and shrunk in the second direction. 2. The method of claim 1, wherein the vertical conductive feature patterns include via patterns. 3. The method of claim 1, wherein the vertical conductive feature patterns include contact patterns. 4. The method of claim 1 further comprising saving the revised vertical conductive feature pattern in the database. 5. The method of claim 1 further comprising forming a photo mask after the step of replacing the identified vertical conductive feature pattern with the revised vertical conductive feature pattern. 6. The method of claim 5 further comprising:forming a first lower-level conductive line and a second lower-level conductive line;forming a dielectric layer on the first and the second lower-level conductive lines;forming a photosensitive layer over the dielectric layer;placing the photo mask on the photosensitive layer;exposing the photosensitive layer;patterning the photosensitive layer;etching the dielectric layer and forming openings; andfilling the openings with a conductive material. 7. The method of claim 1, wherein the method is performed by a computer aided design utility tool. 8. The method of claim 1, wherein the revised vertical conductive feature pattern has a size substantially close to a size of the respective identified vertical conductive feature pattern. 9. The method of claim 1, wherein the revised vertical conductive feature pattern has a size substantially greater than a size of the respective identified vertical conductive feature pattern. 10. The method of claim 1, wherein the first direction is in a direction parallel to a longitudinal direction of a metal line connecting to the respective identified vertical conductive feature pattern. 11. A method of forming a photo mask used in integrated circuit fabrication processes, the method comprising:determining a minimum spacing;identifying patterns having a spacing less than the minimum spacing, wherein the patterns are for forming vertical conductive features;for each identified pattern, determining a first direction to expand and a second direction to shrink to define a revised pattern;checking against design rules to see if the design rules are violated if the respective identified pattern is expanded in the first direction and shrunk in the second direction;replacing each identified pattern with the respective revised pattern when the design rules are not violated, wherein the revised pattern is expanded in the first direction and shrunk in the second direction; andforming a photo mask using the revised pattern. |
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055235140 | claims | 1. Process for the treatment of solid products or liquid products containing plutonium and/or other radioactive elements, comprising the steps of: a) introducing the solid or liquid products into an aqueous, acid solution that is free of acids containing more than one fluorine moiety per molecule, said products comprising organic or mineral waste materials contaminated by plutonium, metallic plutonium or plutonium oxide with said solid products including plutonium or plutonium dioxide, and b) microwave heating the aqueous, acid solution containing these products at a temperature and for a time adequate for destroying the solid products and for dissolving the plutonium and/or radioactive elements in the aqueous solution, said aqueous solution being a solution containing 5 to 14 mole/l HNO.sub.3 and 0.01 to 0.1 mole/l of HF. a) introducing the waste products into an aqueous, acid solution that is free of acids containing more than one fluorine moiety per molecule, said waste products being contaminated by plutonium, metallic plutonium or plutonium oxide, said waste products including at least one contaminated organic solvent, and b) microwave heating the aqueous, acid solution containing these products at a temperature and for a time adequate for destroying the waste products and for dissolving the plutonium and/or radioactive elements in the aqueous solution. 2. Process for the treatment of waste products containing plutonium and/or other radioactive elemental, comprising the steps of: 3. Process according to claim 2, wherein the aqueous solution is a nitric acid solution. |
description | The present application relates generally to nuclear reactors; and more particularly to, a system for dampening the level of vibration experienced by system piping within a nuclear reactor pressure vessel. One type of nuclear reactor, a conventional boiling water reactor (BWR) is shown in FIG. 1. During operation of the reactor, coolant water circulating inside a reactor pressure vessel (RPV) 10 is heated by nuclear fission produced in the nuclear fuel core 35. Feedwater is admitted into the RPV 10 via a inlet 15 and a sparger pipe 20, which is adjacent a core spray line 105. The flows downwardly through a downcomer annulus 25, which is an annular region between RPV 10 and a core shroud 30. The core shroud 30 is a stainless steel cylinder that surrounds the nuclear fuel core 35, which includes a plurality of fuel bundle assemblies 40 (only a few are illustrated in FIG. 1). A top guide 45 and a core plate 50 supports each of the fuel bundle assemblies 40. The coolant water flows downward through the downcomer annulus 25 and into the core lower plenum 55. Then the water in the core lower plenum 55 flows upward through the nuclear fuel core 35. In particular, water enters the fuel bundle assemblies 40, wherein a boiling boundary layer is established. A mixture of water and steam exits the nuclear fuel core 35 and enters the core upper plenum 60 under the shroud head 65. The steam-water mixture then flows through standpipes 70 on top of the shroud head 65 and enters the steam separators 75, which separate water from steam. The separated water is recirculated back to the downcomer annulus 25 and the steam flows out of the RPV 10 and to a steam turbine, or the like, (not illustrated). The BWR also includes a coolant recirculation system, which provides the forced convection flow through the nuclear fuel core 35 necessary to attain the required power density. A portion of the water is drawn from the lower end of the downcomer annulus 25 via recirculation water outlet 80 and forced by a recirculation pump (not illustrated) into a plurality of jet pump assemblies 85 (one is illustrated in FIG. 1) via recirculation water inlets 90. The jet pump assemblies 85 are typically circumferentially distributed around the core shroud 30 and provide the required reactor core flow. A typical BWR has sixteen to twenty-four inlet mixers 95. As illustrated in FIG. 1, a conventional jet pump assembly 85 comprises a pair of inlet mixers 95. Each inlet mixer 95 has an elbow welded thereto, which receives pressurized driving water from a recirculation pump (not illustrated) via an inlet riser 100. A type of inlet mixer 95 comprises a set of five nozzles circumferentially distributed at equal angles about the inlet mixer axis (not illustrated in the Figures). Here, each nozzle is tapered radially inwardly at the nozzle outlet. This convergent nozzle energizes the jet pump assembly 85. A secondary inlet opening (not illustrated) is radially outside of the nozzle exits. Therefore, as jets of water exit the nozzles, water from the downcomer annulus 25 is drawn into the inlet mixer 95 via the secondary inlet opening, where mixing with water from the recirculation pump then occurs. During RPV 10 operation, the flow through the sparger pipe 20 contains pressure fluctuations from various sources in the reactor system. These pressure fluctuations can have frequencies close to one or more natural vibration modes of the sparger pipe 20. The vibration modes experienced by the sparger pipe 20 depends, in part on, on preload and welds on the lugs 110 (not illustrated in FIG. 1) of the sparger pipe 20. The welds may also serve to keep the sparger pipe 20 at a specific distance away from the core spray line 105. In addition to pressure fluctuations, there may be other sources of vibration that can have frequencies close to one or more natural vibration modes of the sparger pipe 20. When an excitation frequency is near the natural frequencies of the sparger pipe 20, the resulting vibration may crack the welds on the lugs 110 and remove the preload from the sparger pipe 20. This can result in loss of the indication of core flow, which may require shutdown of the RPV 10. There are a few possible problems with the currently known systems for dampening the vibration. Currently known systems involve re-welding the lugs 110, which may lead to a repeat failure. These systems generally require longer installation time and expose operators to longer period of radioactivity. For the aforementioned reasons, there is a need for a new system for dampening the vibration experienced by the sparger pipe 20. The system should not require welding. The system should reduce the installation time and lower operator exposure to radioactivity. In accordance with an embodiment of the present invention, an apparatus for dampening vibration experienced by an object integrated with a structure within a reactor pressure vessel (RPV) of a nuclear power plant; wherein the apparatus comprises: a bearing plate configured for providing a barrier between an object and a structure, wherein the structure is located within a reactor pressure vessel (RPV) of a nuclear power plant; a lower section for holding a portion of the object, wherein a first surface of the lower section integrates with the bearing plate and a second surface of the lower section holds a portion of the object; an upper section for holding another portion of the object, wherein the upper section integrates with the bearing plate and mates with the lower section; wherein the lower section and the upper section cooperatively secure the object at a distance from a facing surface of the bearing plate, and allows for dampening of a vibration experienced by the object. In accordance with another embodiment of the present invention, a system for reducing the vibration experienced by a pipe within a nuclear reactor pressure vessel (RPV), the system comprising: a nuclear fuel core comprising a plurality of fuel bundle assemblies; a inlet; a sparger pipe; a core spray line; and a clamp comprising an upper section and a lower section; wherein the clamp is connected to a portion of the sparger pipe to reduce a level of vibration experience by the sparger pipe; and wherein the clamp applies a compressive load to the sparger pipe and positions the sparger pipe and a distance from the core spray line. Certain terminology is used herein for convenience only and is not to be taken as a limitation on the invention. For example, words such as “upper,” “lower,” “left,” “front”, “right,” “horizontal,” “vertical,” “upstream,” “downstream,” “fore”, and “aft” merely describe the configuration shown in the Figures. Indeed, the components may be oriented in any direction and the terminology, therefore, should be understood as encompassing such variations unless specified otherwise. Furthermore, the following discussion focuses on an embodiment of the present invention integrated with the sparger system of the RPV 10. Other embodiments of the present invention may be integrated with other systems that require a dampening of and/or frequency change in vibration. The present invention has the technical effect of reducing the level of vibration experienced by a line, such as, but not limiting of, a pipe, a cable, tubing, or the like, that is connected to at least one separate structure. For example, but not limiting of, the structure includes: a RPV 10, a sparger pipe 20, steam generator, a pipe, a pressure vessel, a heat exchanger, a pump, a condenser, a tank, or the like. An embodiment of the present invention may provide support and a preload to the line at a new location or may replace an existing support, such as, but not limiting of, a weld; which may alter the natural frequencies to avoid resonance from occurring when the structure(s) is excited. An embodiment of the present invention takes the form of an apparatus or system that may reduce the level of vibration experienced by a sparger pipe 20 or other similar object within a RPV 10. An embodiment of the present invention provides at least one repair clamp that generally adds support, to the sparger pipe 20. After installation, the repair clamp may lower the amplitude of, and/or change the frequency of, the vibration experienced by the sparger pipe 20. Referring again to the Figures, where the various numbers represent like parts throughout the several views. FIG. 2 is a schematic illustrating a partially exploded isometric view of an embodiment of a repair clamp 200 within an environment in the present invention operates. FIG. 2 illustrates the repair clamp 200 partially surrounding a portion of the sparger pipe 20. Here, the sparger pipe 20 includes a plurality of lugs 110. An embodiment of the repair clamp 200 may comprise three main components. An upper section 205 for restraining a portion of the sparger pipe 20. A lower section 210 for restraining another portion of the sparger pipe 20 and for receiving a portion of the upper section 205. A bearing plate 215 for providing a barrier between the sparger pipe 20 and a wall of the RPV 10. Generally, the upper section 205 and the lower section 210 cooperatively operate to secure the sparger pipe 20 at a desired distance from the wall of the RPV 10 and from a surface of the core spray line 105. When installed the repair clamp 200 may add or restore a preload to the sparger pipe 20, while changing the amplitude and/or frequency of the potential vibration. FIGS. 3A-3B, collectively FIG. 3, are schematics illustrating exploded isometric views of a repair clamp 200 in accordance with an embodiment of the present invention. FIG. 3A illustrates embodiments of the upper section 205 and the bearing plate 215 with the present invention. FIG. 3B illustrates an embodiment of the lower section 210 in with present invention. Referring now to FIG. 3A, an embodiment of the upper section 205 may comprise: a second crimp collar 230, a second jacking bolt 235, a pinch plate 240, and at least one pinch bolt 245, all of which may assemble onto a upper clamp body 250 of the upper section 205 as illustrated, for example, but not limiting of, in FIG. 3A. The second jacking bolt 235 serves to connect the upper clamp body 250 with the hearing plate 215. The second jacking bolt 235 may comprise a variety of forms. An embodiment of the second jacking bolt 235 may comprise a cylindrical shaft with a series of threads, or other groves, that allow for mating with the second crimp collar 230 and the upper clamp body 250. An end portion of the second jacking bolt 235 may allow for a hand tool to rotate the second jacking bolt 235. For example, but not limiting of, the end portion may allow for a wrench to fasten the second jacking bolt 235 to the upper clamp body 250. The second crimp collar 230 serves to fix the second jacking bolt 235 to a desired position. This may prevent the second jacking bolt 235 from loosening due to vibration after the repair clamp 200 is installed. An embodiment of the second crimp collar 230 may allow for the second jacking bolt 235 to pass through and mate with a portion of the upper clamp body 250, as illustrated, for example, but not limiting of, in FIG. 3A. An embodiment of the second crimp collar 230 may also allow for the at least one pinch bolt 245 to pass through. Here, the second crimp collar 230 may prevent the at least one pinch bolt 245 from loosening, due to vibration, after the repair clamp 200 is installed. The at least one pinch bolt 245 serves to apply a compressive load from the repair clamp 200 to the sparger pipe 20. Essentially, the at least one pinch bolt 245 clamps the upper section 205 and the lower section 210 around a portion of the sparger pipe 20. An embodiment of the upper clamp body 250 may allow for the lug 110 to mate with a surface on the upper clamp body 250. This may allow for a robust connection between the repair clamp 200 and the sparger pipe 20. The pinch plate 240 may serve to provide a bearing surface between the at least one pinch bolt 245 and the second jacking bolt 235. As illustrated in FIG. 3A, the upper clamp body 250 may comprise a plurality of recesses allowing for the aforementioned upper section 205 components to mate therein. The upper section 205 may also comprise groves for a top end of the bearing plate 215 to slidably attach. The upper clamp body 250 may also comprise a tang portion that allows for mating with the lower section 210, as described below. A surface of the upper clamp body 250 may be of a shape allowing for mating with the portion of the sparger pipe 20. For example, but not limiting of, the shape may comprise an arc of a similar radius of the outer diameter of the sparger pipe 20. Moreover, the surface may comprise a notch, or the like, that allows for mating with the lug 110. The bearing plate 215 serves as a barrier between the repair clamp 200 and a wall of the RPV 10. Generally, the bearing plate 215 may form the rear of the repair clamp 200 encompassing the rear portions of the upper section 205 and the lower section 210. In an embodiment of the present invention an overall length of the bearing plate 215 extends beyond an overall length of the upper section 205 mated with the lower section 210. This feature may allow for the applied forces of the second jacking bolt 235 and the first jacking bolt 225 to be transferred on the wall of the RPV 10. Referring now to FIG. 3B that illustrates an exploded view of the lower section 210. An embodiment of the lower section 210 may comprise a first crimp collar 220 and a first jacking bolt 225, both of which may be integrated with a lower clamp body 227. The first crimp collar 220 generally functions similar to that of the second crimp collar 230, as previously described. The first jacking bolt 225 generally functions similar to that of the second jacking bolt 235, as previously described. The lower clamp body 227 serves to receive and hold a portion of the sparger pipe 20. A first surface of the lower clamp body 227 may integrate with the bearing plate 215. A second surface of the lower clamp body 227 may be of a shape allowing for mating with the portion of the sparger pipe 20. For example, but not limiting of, the shape may comprise an arc of a similar radius of the outer diameter of the sparger pipe 20. Moreover, the second surface may comprise a notch, or the like, that allows for mating with the lug 110. FIGS. 4A-4B, collectively FIG. 4, are schematics illustrating isometric and side elevation views of an assembled repair clamp 200 securing a portion of a sparger pipe 20 in accordance with an embodiment of the present invention. FIG. 4A illustrates a side isometric view. FIG. 4B illustrates an opposing side elevation view. FIG. 4A specifically illustrates the upper section 205 and lower section 210 assembled and clamping onto the sparger pipe 20. Here, FIG. 4A illustrates how the tang portion of the upper clamp body 250 may slidably mate with the lower clamp body 227 in accordance with an embodiment of the present invention. FIG. 4A also illustrates how the ends of the first jacking bolt 225 and the second jacking bolt 235 may extend through holes mating holes. Specifically, the upper clamp body 250 may have a hole that allows the second jacking bolt 235 to extend through and engage the bearing plate 215. Also, the lower clamp body 227 may comprise a recess or a notch that allows the tang portion of the upper clamp body 250 to fit therein. This may allow for the tang and a rear portion of the lower clamp body 227 to form the relatively flat rear portion of the lower section 210. Here, an embodiment of the upper clamp body 250 may comprise the tang portion having a hole that allows for the first jacking bolt 225 to extend through to engage the bearing plate 215. FIG. 4A also illustrates how the position at least one pinch bolt 245 may be maintained by the second crimp collar 230. Here, after the at least one pinch bolt 245 provides the compressive load to secure the sparger pipe 20, the second crimp collar 230 may be used to prevent loosening, as previously described. Referring now to FIG. 4B, which is an opposing side elevation view of FIG. 4A. FIG. 4B illustrates the repair clamp 200 having the bearing plate 215 and the pinch plate 240 connected to the upper section 205. As discussed, in an embodiment of the present invention the bearing plate 215 may serve as the back portion of the repair clamp 200. Also, the pinch plate 240 may provide a barrier between the second crimp collar 230 and the bearing plate 215. FIG. 4B also illustrates how the first jacking bolt 225 and the second jacking bolt 235 may reside within the respective crimp collar, 220, 230, as previously described. FIG. 4B also illustrates how the upper section 205 and the lower section 210 may partially surround a portion of the sparger pipe 20. FIG. 5 is a schematic illustrating a plan view elevation of a repair clamp 200 installed on a sparger pipe 20, in accordance with an embodiment of the present invention. FIG. 5 illustrates a few of the benefits of an embodiment of the present invention. A first benefit allows for a generous clearance around the repair clamp 200. When fully engaged the repair clamp 200 may provide a vertical distance of between the sparger pipe 20 and the core spray line 105, represented by dimension “X” in FIG. 5. When fully engaged the repair clamp 200 may also provide a distance of between a bottom surface of the repair clamp 200 and the core spray line 105, represented by dimension “Y” in FIG. 5. When fully engaged the repair clamp 200 may also provide a distance of between a side of the repair clamp 200 and a portion of the sparger pipe 20 represented by dimension “Z” in FIG. 5. The dimensional values and/or ranges of X, Y, and Z may be determined by the type of RPV 10 to which an embodiment of the repair clamp 200 is installed. A second benefit of the repair clamp 200 allows for a range of approximately 0.250 inches of adjustment. This may accommodate a broad range of sparger pipe 20 positions relative to a wall of the RPV 10. A third benefit involves the crimp collars 220,230, which may serves as a positive anti-rotation device. The crimp collars 220,230 may only require a visual inspection to confirm that the repair clamp 200 has not loosened. The components of an embodiment present invention may be formed of any material capable of withstanding the operating environment to which the repair clamp 200 may be exposed. In use, the repair clamp 200 may clamp around the sparger pipe 20 at a location of previous jacking bolts. When fully engaged, the repair clamp 200 may provide for generous clearance around the upper bounds of the sparger pipe 20 tolerance and connecting welds. The repair clamp 200 may restore the preload on the sparger pipe 20 at the location of the previous jacking bolts. The repair clamp 200 may also reduce the vibration experience by the sparger pipe 20. Although the present invention has been shown and described in considerable detail with respect to only a few exemplary embodiments thereof, it should be understood by those skilled in the art that we do not intend to limit the invention to the embodiments since various modifications, omissions and additions may be made to the disclosed embodiments without materially departing from the novel teachings and advantages of the invention, particularly in light of the foregoing teachings. Accordingly, we intend to cover all such modifications, omission, additions and equivalents as may be included within the spirit and scope of the invention as defined by the following claims. For example, but not limiting of, an embodiment of the present invention may be used to: a) introduce a different vibration mode; h) to secure a pipe, cable, wire, or other similar object, at a fixed distance away from a separate structure or other object; or c) to apply a compressive load to at least one of the aforementioned objects. |
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description | This application claims priority of German Application No. 10 2006 017 904.8, filed Apr. 13, 2006, the complete disclosure of which is hereby incorporated by reference. a) Field of the Invention The invention is directed to an arrangement for generating extreme ultraviolet radiation from a plasma generated by an energy beam with high conversion efficiency in which a pulsed energy beam is directed in a plasma generation chamber to a location where it interacts with a target, a target feed device contains a mixing chamber for generating a mixture of particles of an emission-efficient target material with at least one carrier gas and an injection unit for dispensing individually defined target volumes into the plasma generation chamber in a metered manner in order to supply only as much emission-efficient target material to the interaction location as can be converted into radiation by an energy pulse. The invention is applied in particular in radiation sources for EUV lithography for the fabrication of semiconductor chips. b) Description of the Related Art Known “clean fuels” (target materials such as xenon) are not sufficiently efficient for the generation of EUV radiation based on a plasma which is excited by a pulsed energy beam for emitting in the EUV spectral band around 13.5 nm because their conversion efficiency (ratio of the emitted energy in the desired EUV spectral band to the (laser) excitation energy) is only about 1%. By “clean fuel” is meant that it does not produce a “coating” of components of the radiation source, i.e., it does not generate precipitation (contamination) on surfaces (particularly optical surfaces). Metallic target materials (e.g., elements of groups IV to VII of the 5th period of the periodic table of elements) are substantially more efficient for generating EUV at 13.5 nm (e.g., tin has a conversion factor of approximately 3%), but produce a “coating”, i.e., in exciting plasma they generate debris which results especially in precipitation but also leads to ablation of components of the radiation source, especially optical components. Further, ablation processes (removal of material from optical surfaces) which are caused by the high kinetic energy of unconsumed target particles not converted into luminous plasma are appreciably reduced for “clean fuels” (e.g., xenon) compared to metallic target materials. Pure tin (Sn) delivers a broad-band spectrum around 13.5 nm±2% (desired EUV spectral band for semiconductor lithography, so-called EUV in-band radiation) but also has significant proportions outside the desired EUV spectral band for semiconductor lithography (EUV out-of-band radiation). These out-of-band radiation components are undesirable because they contribute to unnecessary heating of the optics and other source components. In order to make use of metal-containing targets, it was known in the prior art to use metallic solutions at room temperature as target droplets for laser-generated punctiform plasma. In U.S. Pat. No. 6,831,963 B2, copper compounds and zinc compounds in particular such as chloride solutions, bromide solutions, sulfate solutions, nitrate solutions and organometallic solutions are described as metallic solutions which can be applied in the vicinity of optical components without damage to the latter because hardly any debris is produced. However, substantially only radiation in the range from 11.7 nm to 13 nm is generated, which must be classified as out-of-band radiation components within the meaning of the above-stated requirements of EUV lithography. The same situation is also described for tin compounds, particularly tin chloride, in US 2004/0208286 A1. As is disclosed in WO 2002/046839 A, an injection of droplets in liquids (e.g., tin as compound or nanoparticle) makes it possible to limit the amount of convertible target material. However, it is disadvantageous that all of the carrier liquids or solvents known for this purpose contain component parts which are damaging to optics (carbon coating, oxygen oxidation, etc.). WO 2004/056158 A2 describes a device for generating x-ray radiation and EUV radiation in which a mist with an atomic density of >108 atoms/cm3 is generated for increasing the target density of the smallest possible droplets (on the order of the laser wavelength). The improved target density is generated by the absorption of the target liquid in a nonreactive gas in that an electro-magnetically switchable valve is connected to an ultrasonic nozzle via an expansion duct which is outfitted with heating means for increasing temperature in order to generate a supersaturated vapor and supply it by bursts through the target nozzle for generating plasma. The disadvantage here consists in the elaborate metering procedure and in that the target density drops off quickly after exiting the target nozzle. Gaseous injections of nanoparticles into a carrier gas, as is described in EP 0 858 249 B1 and WO 2004/084592 A2, are generally not sufficiently concentrated because the particle-containing “gas cloud” expands rather quickly so that the density is too low for an efficient excitation, e.g., by means of a laser, even at a short distance from the injection site (on the order of 1 cm). Therefore, the excitation must be carried out in the vicinity of the injection opening, and limiting the particle quantity to the amount needed for complete energy conversion cannot be accomplished in a simple manner. WO 2004/084592 A2 discloses a possibility for metering solid target material. A chamber system is provided in which a mixing of solid or liquid target clusters in a gas is carried out in a first chamber. As a result, a “focused mass flow” is generated in a second chamber and arrives in the third chamber for plasma generation through a periodically opening shutter device as a pulsed mass flow in order to provide the necessary amount of convertible target material for each laser pulse and accordingly to reduce the proportion of unconverted target material in the plasma chamber. The target material that is blocked in the second chamber by the shutter device is sucked out and can be reused. It is the primary object of the invention to find a novel possibility for generating EUV radiation by means of a plasma induced by an energy beam that pen-nits a more efficient conversion of the energy radiation into EUV radiation in the wavelength region of 13.5 nm by using metallic target material without the optical components arranged downstream being damaged by debris that is generated as a result of excess target material. Further, the target material can be supplied in such a way that radiation is generated at a great distance from the injection device so as to ensure a long lifetime of the injection device. Another object of the invention is to find a form of injection for metallic target material which (a) is suitable for efficient absorption of laser radiation of about 1 μm, (b) contributes to the spectral narrowing of the emission band at 13.5 nm, and (c) does not contain any components apart from the metallic target components that damage the source components essential to operation. In an arrangement for generating extreme ultraviolet radiation from a plasma generated by an energy beam with high conversion efficiency in which a pulsed energy beam is directed in a plasma generation chamber to a location where it interacts with a target, containing a target feed device, a mixing chamber for generating a mixture of particles of an emission-efficient target material with at least one carrier gas, and an injection unit for dispensing individually defined target volumes into the plasma generation chamber in a metered manner in order to supply only as much emission-efficient target material to the interaction location as can be converted into radiation by an energy pulse, the above-stated object is met in that the target feed device has a gas liquefaction chamber, wherein the target material is supplied to the injection unit as a mixture of solid metal particles in liquefied carrier gas, and in that the injection unit has a droplet generator with a nozzle chamber and a target nozzle for generating a defined droplet size and series of droplets, wherein means which are controllable in a frequency-dependent manner and which are triggered by the pulse frequency of the energy beam are connected to the injection unit for generating a time-controlled series of droplets. The liquefaction chamber is advantageously arranged downstream of the mixing chamber so that the solid particles are supplied to the liquefaction chamber so as to be mixed with the carrier gas, and the liquefaction chamber is designed for the liquefaction of the particle-gas mixture. In another advisable variant, the liquefaction chamber is arranged upstream of the mixing chamber so that the liquefaction chamber is designed for the liquefaction of the pure carrier gas, and the mixing chamber is designed for mixing the solid particles with the liquefied carrier gas. The solid emission-efficient particles advantageously comprise tin, a tin compound, lithium, or a lithium compound. The solid particles preferably have a size of less than 10 μm, preferably in the nanometer range and, without limiting generality, are referred to hereinafter as nanoparticles. Inert gases such as nitrogen or noble gases are advantageously used as carrier gas. Argon is very well-suited for this purpose. In addition, light noble gases (e.g., helium, neon) are advisably mixed in with a carrier gas of the type mentioned above as main component in order to limit the spectral band width of the EUV emission at 13.5 nm, i.e., in order to suppress out-of-band radiation. The individual targets (droplets) ejected from the injection unit advantageously have a diameter between 0.01 mm and 0.5 mm. It has proven particularly advantageous for reducing the contamination caused by excess target material when means for removing individual targets are arranged downstream of the target nozzle of the injection unit so that the frequency of the individual targets arriving in the interaction location exactly corresponds to the pulse frequency of the energy beam. In an advantageous first variant, electric or magnetic deflecting means are arranged downstream of the target nozzle of the injection unit for selective lateral deflection of unnecessary individual targets from the series of droplets dispensed by the target nozzle. In a second construction for eliminating individual targets, a mechanical closure device (e.g., a mechanical shutter, chopper wheel) is provided after the target nozzle of the injection unit for defined elimination or passage of individual targets from the series of droplets dispensed by the target nozzle. In a third variant, the injection unit has a target generator with a pressure modulator at the nozzle chamber in order to increase the chamber pressure temporarily for ejecting an individual droplet when needed and has a nozzle antechamber which is arranged downstream of the target nozzle and in which a pressure is maintained that is higher than that of the plasma generation chamber and adapted to the gas pressure of the gas feed to the mixing chamber. Adapting the pressure in the nozzle antechamber surrounding the target nozzle prevents unwanted dripping of target material from the target nozzle as long as no pressure pulse is generated by the pressure modulator. For a suitable pressure adaptation in the nozzle antechamber, the pressure of the gas feed to the mixing chamber is preferably adjusted so as to be slightly higher (on the order of 0.5 to 1 bar higher) than that in the nozzle antechamber. For producing the liquid particle-gas mixture, a sufficient quantity of particles can also advisably be provided in a reservoir and supplied to a plurality of mixing chambers which are arranged in parallel and connected to the target generator so as to be switchable in series for continuous injection into the plasma generation chamber. In another advantageous variant, the particles are provided so as to be mixed with the carrier gas in a mixing chamber and a line connection point with a feed line from another carrier gas feed is arranged downstream of the mixing chamber, and at least one of the feed lines to the connection point has a throughflow regulator which is controlled by a measuring device which is arranged downstream of the connection point and which determines the proportion of particles in the gas flow in order to adjust a desired mixture ratio of mixed carrier gas and pure carrier gas. The measuring device for controlling the mixture ratio is preferably an optical scatter light measuring unit. The pulsed energy beam needed for plasma excitation can comprise at least one laser beam, an electron beam, or an ion beam. The fundamental idea of the invention is based on the consideration that the conversion of radiated excitation energy into the desired radiation band of 13.5 nm by the excitation of metallic target materials, particularly tin, with a pulsed energy beam is very efficient (three times the conversion efficiency of xenon which is conventionally used). However, metals can be used in a radiation source for EUV lithography only by ensuring extensive absence of contamination which, as is well known, can be achieved by limiting the emitting target material to the amount needed for generating radiation. The invention solves this problem through the combination of generating a mixture of solid metal particles (nanoparticles with diameters <10 μm) with an inert carrier gas, gas liquefaction, and a metered injection of droplets into the plasma generation chamber. Supplying the liquid mixture of solid metal particles and carrier gas to the plasma generation chamber by means of an injection device in the form of a droplet generator makes possible (compared to gas puffs) a substantially higher target density and an appreciably greater distance between the location of interaction of the target with the energy beam and the injection location so that radiation yields (conversion efficiency) and contamination (damage to the injection nozzle by debris) are considerably reduced. When noble gases or nitrogen which themselves do not contain optics-damaging components are used as carrier medium, the liquid target material generated in this way does not lead to further contamination. Sn nanoparticles are preferably used as emitters and, by mixing in a light carrier gas (helium and/or neon) with the main carrier gas, unwanted spectral bands outside the EUV band for semiconductor lithography are extensively suppressed. Liquefied noble gas or liquid nitrogen can also be used directly for the particle mixture. The inventive solution makes it possible to generate EUV radiation by means of a plasma induced by an energy beam, which permits a more efficient conversion of the energy radiation into EUV radiation in the wavelength region of 13.5 nm without optical components arranged downstream being further damaged by excess target material. Further, the great distance that can be achieved between the plasma and the injection device ensures a longer life of the injection device and a more stable generation of radiation. The invention will be described more fully in the following with reference to embodiment examples. The EUV radiation source has a target feed device 1 which, as is shown schematically in FIG. 1, basically contains a mixing chamber 11, a liquefaction chamber 12 and an injection unit 13. The injection unit 13 has a droplet generator 131, a pressure modulator 132, a target nozzle 133, and a nozzle chamber 134. Solid particles 14 comprising metals or metal compounds, e.g., tin or lithium (or preferably also their oxides, SnO, SnO2, LiO, LiO2) which emit efficiently in the EUV spectral region (around 13.5 nm) and a clean (i.e., free from emitting particles) carrier gas 15, e.g., noble gases or nitrogen, are combined and mixed in the mixing chamber 11. The resulting particle-containing mixture 16 is fed to the liquefaction chamber 12, wherein liquefaction is carried out at low temperatures (T<173 K) and pressures >1 bar. Sn particles (individual particles of at most 10 μm in size) are preferably mixed in to achieve a high efficiency of EUV generation (≈3%). However, mixtures of other elements (e.g., lithium) or compounds (preferably tin compounds or lithium compounds) are also possible. As is shown schematically in FIG. 1, the mixture of the particles 14 with the carrier gas 15 in a gas phase is carried out in that the particles 14 and the carrier gas 15 are combined in a mixing chamber 11. A number of methods for isolating particles from an existing bulk mass and introducing them into a gas flow in a metered manner are known from particle technology. One possible method is to pull the particles individually out of the bulk mass by means of a special rotating brush and transfer them to a carrier gas flowing past the brush. But the particles 14 can also be present in sufficient quantity in a mixing chamber 11 and, for continuous operation of the EUV source, switching is carried out between a plurality of mixing chambers 11 which are connected in parallel. It is also possible to mix the solid particles 14 into an already existing liquid gas 17 as will be described more fully in the example referring to FIG. 5. The particle-containing liquid gas 17 is supplied to the injection unit 13 and introduced into the nozzle chamber 134. A stable continuous series 2 of droplets is dispensed along a target axis 21 in the plasma generation chamber 3 by means of a pressure modulator 132 (e.g., piezo-actuator) via the target nozzle 133 in tune with the drop breakup frequency of the liquid gas 17. An energy beam 4 is directed to the target axis 21 at the desired interaction location 41, and the successive pulses of this energy beam 4 respectively excite an individual target 23 (droplet) to form EUV-emitting plasma 5 when this individual target 23 passes the interaction location 41. The target feed device 1 is incorporated together with the housing of the injection unit 13 in the plasma generation chamber 3. The housing of the injection unit 13 forms a nozzle antechamber 135 around the target nozzle 133 in order to adjust a higher pressure relative to the evacuated plasma generation chamber 3 so that the exit of liquid gas and the droplet formation are stabilized. The target feed device 1 can also be introduced into the plasma generation chamber 3 at other positions, e.g., at the feed line between the liquefaction chamber 12 and the injection unit 13 or between the mixing chamber 11 and the liquefaction chamber 12. According to FIG. 1, without limiting generality, a series 2 of droplets of the individual target 23 is generated in tune with the natural drop breakup frequency in that a closed target jet 22 is initially generated which passes into a stable, continuous series of individual targets (droplets) 23 shortly after exiting the target nozzle 133. In general, as is shown schematically in FIG. 1, not every individual target 23 can be struck by a pulse of the energy beam 4. However, droplets 23 which fly past the interaction location without being used can be sucked out at the end of the target axis 21 virtually without damage in a sink coupled with a vacuum pump (not shown). The injection of the particle-containing liquid gas 17 is carried out in such a way that droplets 23 are formed in the desired size, generally in the form of solid globules, when they reach the interaction location 41 because the liquid gas 17 expands adiabatically and freezes when injected into the vacuum of the plasma generation chamber 2, i.e., after exiting the nozzle antechamber 135 (at higher pressure). The size of the droplets 23 is defined by the amount of mixture that is optimally excited to form a radiating plasma 5 at a given energy of an excitation pulse of the energy beam 4. The proportion of solid particles 14 in the liquid gas 17 is adjusted in such a way that the efficiency of the EUV generation and the width of the spectrum are optimized. In this way, a limiting of the amount of the Sn particles 14 assumed herein is achieved, i.e., the amount of Sn in the plasma generation chamber 3 is limited to the amount needed for generating radiation so that no excess metallic target material which, as debris, could damage the components of the radiation source as a result of insufficient excitation, remains in the plasma generation chamber 3. The carrier gas 15 (N2 or a noble gas) can at most be potentially damaging to the optics due to the kinetic energy of its particles. A suppression of sputter processes of this kind is easily possible and is known from xenon-based EUV sources, e.g., by means of introducing a blocking gas (e.g., argon cross-flow) between the plasma 5 and the collector optics. In any case, the carrier gas 15 itself does not contain any component parts that are damaging to optics such as carbon (C) or oxygen (O2). Because of the injection of the particle-containing mixture 16 in liquid form, a very great distance can be achieved between the generation of radiation (plasma 5) and all of the important components of the system such as the target nozzle 133, collector optics for bundling the generated EUV radiation (not shown), etc. The large distance results in a longer life of these components. In particular, the target nozzle 133 is also substantially less damaged (eroded) by heat radiation and particle radiation from the plasma 5 so that a stable target supply in the interaction location 41 can be achieved over a longer operating period. Because of the coating property of metallic “fuels” (solid targets), their amount must be limited to the amount necessary for generating radiation. When using tin (Sn), which has strong spectral lines at 13.5 nm, about 5·1014 Sn ions (this corresponds to an Sn volume of about 30 μm diameter) are required for an EUV source size of 0.5 mm diameter with an excitation energy of about 1 J per individual excitation. The source size is derived from the etendue requirement of EUV lithography. The small Sn volume can reasonably be adapted in size to the required source size of the emission prior to excitation by expansion with a pre-pulse of the energy beam 4. The necessary energy is on the order of 10 mJ and is carried out approximately 100 ns before introducing the high-energy pulse. At a repetition frequency of about 10 kHz, a source with these parameters behind collector optics would reach an EUV in-band output (13.5 nm±2%) of about 100 W. The Sn consumption per day in this case is about 85 g when the quantity of Sn is limited to the amount needed for generating radiation. The ion density (and electron density) is derived solely from the optimized EUV emission for a homogeneous volume. The electron density is too low for efficient absorption of laser radiation with a wavelength of 1 μm. Therefore, the carrier gas 15 functions additionally as an electron donor to achieve a laser absorption of almost 100%. This is ensured for nitrogen (N2) and argon (Ar) in a stoichiometric proportion of the carrier gas from about 2/3. The stoichiometric proportion is the ratio of the quantity of atoms or molecules of target material (bound in particles) and carrier gas in relation to a volume element. In addition, by mixing in lighter carrier gases (He, Ne) the spectral bandwidth of the radiation emission of tin at 13.5 nm is reduced, whereas with pure tin it is appreciably greater than the required ±2% (J. Opt. Soc. Am. B 17 (2000) 1616, Choi et al.). Further, the proportion of radiation outside the desired EUV spectrum is likewise appreciably reduced. A true limiting of the amount of “fuel” (solid particles 14) to the amount needed for generating radiation is only achieved when the target volumes are supplied at a frequency that exactly matches the frequency at which the energy pulses are introduced (on the order of 10 kHz), i.e., exactly one target volume is supplied to the interaction location 41 for each individual generation of radiation. In the following three examples, compared to a variant shown in FIG. 1, to generate a particle-containing series 2 of droplets at high frequency (typically 100 kHz), wherein the natural drop breakup frequency is stabilized by a pressure modulator 132, individual volumes are removed (by various steps) from the series 2 of droplets which is generated at too great a density, so that as a result the frequency of the volumes in the interaction location 41 (plasma 5) matches the frequency of the energy pulses. FIG. 2 shows an EUV source constructed in the above manner in which it is assumed without limiting generality that the energy beam 4 is a laser beam 42. The target feed device 1 differs from that shown in FIG. 1 in that an electric deflecting device 136 and a suction device 137 are connected to the injection unit 13 downstream of the output of the nozzle antechamber 135 in order to “thin” the dense series of droplets 23 and adapt the frequency of the droplets 23 in the location 41 of interaction with a laser beam 42 exactly to the pulse repetition frequency of the laser. The excess droplets 23 are removed by the suction device 137 and supplied again to the liquefaction chamber 12. In this way, in contrast to the construction in FIG. 1, excess droplets 23 are prevented from partially evaporating in the immediate vicinity of the plasma 5 or from contributing generally to the increase in the gas load inside the plasma generation chamber 3. In a second variant (according to FIG. 3), the particle-containing droplets 23 are already generated so as to correspond exactly to the pulse frequency of the laser beam 42. FIG. 3 shows a modified droplet selection in which pressure compensating means 138 which supply a pressure pantechamber approximately corresponding to the gas pressure pcarrier gas supplied to the mixing chamber 11 are connected directly to the nozzle antechamber 135. Accordingly, the droplets 23 are released through the pressure modulator 132 with exactly the same frequency as the pulse frequency of the laser beam 42 so that the injection device 13 ejects droplets 23 only in such quantity that every droplet 23 is struck by exactly one pulse of the laser beam 42. This is realized in a reliable manner in that the nozzle antechamber 135 of the injection unit 13 downstream of the target nozzle 133 is connected to pressure compensating means 138 which are adapted to the pressure pcarrier gas of the gas feed to the mixing chamber 11 so that the liquid target material cannot form any unwanted droplets 23 in the nozzle chamber 134 and enter the plasma generation chamber 3 without a temporary pressure increase of the pressure modulator 132. The pressure modulator 132 which can be, e.g., a piezo-actuator arranged at the nozzle chamber 134 generates pressure pulses at the frequency of the energy pulses, i.e., only individual targets 23 are supplied as needed (corresponding to the triggered pulses of the laser beam 42). FIG. 4 shows a droplet selection having the same effect as that in FIG. 3 in which exactly one individual droplet 23 is associated with each pulse of the laser beam 42. In this construction, however, mechanical means in the form of a rotating aperture plate 32 are provided to pass only every nth droplet 23 into the plasma generation chamber 3. At the same time, the aperture plate 32 makes up part of a vessel wall which partitions the plasma generation chamber 3 to form an antechamber 31, and a higher pressure pantechamber is adjusted in the antechamber 31 as in the previous examples in the nozzle antechamber 135. Therefore, a separate nozzle antechamber 135 of the injection unit 13 can be dispensed with in this example. It is shown schematically in FIG. 4 that every second droplet 23 is intercepted on the aperture plate 32 and sublimed or evaporated thereon and can be sucked out of the antechamber 31 through a separate pump unit (not shown). Under real conditions, only about every tenth droplet 23 is passed for interaction with the laser beam 42. As was already mentioned above, it is also useful to mix solid particles 14 into carrier gas 15 which has already been liquefied beforehand. An arrangement of this kind is shown in FIG. 5. In this construction, the mixing chamber 11 and the liquefaction chamber 12 are reversed with respect to the preceding examples. Further, the carrier gas is fed into the liquefaction chamber 12, and the liquid gas 17 produced therein is introduced into the mixing chamber 11 so as to be mixed with the solid particles 14. Otherwise, the construction is the same as that shown in FIG. 1, but could also be realized according to the constructions in FIGS. 2 to 4. A preferred variant of the invention is shown in FIG. 6. In this case, it is assumed that the solid emission-efficient particles 14 are already mixed with the carrier gas 15 in a mixing chamber 11 functioning as a reservoir. In order to isolate the particles 14 from the existing bulk mass (not shown) and introduce them into a gas flow in a metered manner, the particles 14 are removed individually from the bulk mass by a rotating brush and are transferred to a flow of carrier gas 15 which flows past. As the flow of gas proceeds, it must be ensured through a suitable design of the lines conducting the carrier gas that the particles do not become unmixed. The line proceeding from the mixing chamber 11 in direction of the injection unit 13 is then tied to another carrier gas line in a connection point (+) in such a way that the gas flows can be regulated relative to one another by means of a throughflow regulator 16 prior to the connection point (+). A measuring device 19 arranged downstream of the connection point (+) serves to determine a regulating variable. The measuring device 19 measures the actual mixture ratio, e.g., by measuring scatter light, and accordingly supplies a correcting variable for the relative adjustment of the supplied amounts of clean carrier gas 15 and particle-containing mixture 16. This additional admixing of carrier gas enables a very accurate adjustment of the proportion of solid particles 14 per volume unit of carrier gas 15 and therefore a highly accurate metering of the effective target quantity (particles 14) per droplet 23 of the liquid gas generated therefrom. Although FIG. 6 shows both feed lines of the clean carrier gas 15 and particle-containing mixture 16 to the connection point (+) with throughflow regulators 18, it would also be sufficient when one of the feed lines, preferably the carrier gas feed line, is outfitted with a throughflow regulator 18. Further, the measuring device 19 which directly influences the pressure adjustment in front of the liquefaction chamber 12 according to FIG. 6 can also be used for an adapted pressure regulation of the pressure pantechamber in the nozzle antechamber 135. Accordingly, the construction shown in FIG. 4 makes possible a suitably adapted pressure regulation for supplying droplets 23 exclusively when needed (drop on demand), i.e., so as to correspond to the pulse rate of the laser beam 42. While the foregoing description and drawings represent the present invention, it will be obvious to those skilled in the art that various changes may be made therein without departing from the true spirit and scope of the present invention. Reference Numbers 1 target feed device 11 mixing chamber 12 liquefaction chamber 13 injection unit 131 droplet generator 132 pressure modulator 133 target nozzle 134 nozzle chamber 135 nozzle antechamber 136 deflecting device 137 suction device 138 pressure compensating means 14 (solid) particles 15 carrier gas 16 particle-containing mixture 17 liquid gas 18 throughflow regulator 19 measuring device 2 series of droplets 21 target axis 22 target jet 23 individual target (droplet) 3 plasma generation chamber 31 antechamber (of the plasma generation chamber) 32 (rotating) aperture plate 4 energy beam 41 interaction location 42 laser beam 5 plasma p pressure |
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042474957 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for the manufacture of UO.sub.2 nuclear fuels containing PuO.sub.2 in set amounts present as a solid solution with the UO.sub.2, and in the reprocessing of such nuclear fuel with nitric acid, the PuO.sub.2 is soluble in the nitric acid solvent. 2. Background of the Invention In addition to uranium 235, the plutonium obtainable from burned-up fuel assemblies is known to be fissionable, also producing energy. Its best utilization is in so-called breeder reactors, in which over a certain period of time (doubling time) more plutonium is produced than the reactor itself consumes. But plutonium as dioxide mixed with uranium dioxide can also be burned up in so-called thermal reactors, producing energy. For this purpose, it is mixed with uranium as oxide in accordance with various methods and also used in various forms as nuclear fuel. In the reprocessing of such nuclear fuels for obtaining the plutonium especially the breeder plutonium, difficulties are encountered, however, if the plutonium dioxide is not present as a solid solution with the uranium dioxide. This difficulty is due to the fact that it is not possible to dissolve as completely as possible the plutonium dioxide in the solvent HNO.sub.3 provided for UO.sub.2 fuel. Substantially complete dissolution of the plutonium dioxide is necessary for the reprocessing. This is possible only if additives, such as fluoric acid or hydrofluoric acid or fluorine ions or other additions containing dissolving ions, are admixed. In view of the corrosive action of the fluorine ion or of fluoric acid or also other dissolving ions on parts of the installation, as well as particularly also in view of the negative effects of adding fluorine ions in the glassification or encapsulation of the waste from the reprocessing, one endeavors to keep them but of the process. A number of processes are known in principle, by which homogeneous solid solutions and soluble plutonium dioxide nuclear fuel can be produced, as for instance, melting the oxides UO.sub.2 and PuO.sub.2 ; the joint precipitation of different or similar compounds of uranium and plutonium; as well as also thorough milling of both oxide components, as described in the German Published Prosecuted Application No. 1 571 343. However, the methods known to date are either too laborious and expensive such as, for instance, jointly melting the uranium dioxide and the plutonium dioxide at above 2000.degree. C., or the method is difficult to control and, in particular, requires the presence of an aqueous solution of the uranium and the plutonium, which is not a simple procedure. Furthermore, the residue insoluble in nitric acid which is left in most of the processes of joint precipitation, as well as in the dry milling processes known so far, is not small enough in all cases to ensure the reprocessing of burned-up fuel with high reliability without the addition of fluorine ions or the use of other measures. The reprocessing aspect, however, is of paramount importance for the future of nuclear reactor technology. Experiments have shown that, for example, in the preparation of mixed oxides of 75% UC.sub.2+x powder and 25% PuO.sub.2 powder, which have been mixed by ball milling for several hours, and after sintering the powder, pressed into pellets, at 1700.degree.C. for four hours, an insoluble plutonium content of about 3%, referred to the plutonium input, was found. On the basis of these findings, it can be assumed as certain that the nuclear fuel prepared in accordance with German Published Prosecuted Application No. 1 571 343, column 3, lines 18 to 23, also does not have sufficient solubility, as there, the milling and sintering is likewise performed only once. As found in the literature (e.g. Report of Oak Ridge National Laboratory No. ORNL/TM-5909), the results of dissolution tests on non-irradiated mixed oxides which were prepared by joint precipitation of ammonium diuranate and plutonium hydroxide and, further, processing into nuclear fuel pellets, scatter from 0.1 to 1% of the undissolved plutonium referred to the plutonium input. In various investigations, it has also been found that a uranium-containing phase is present in the undissolved residue, in addition to a plutonium-containing phase. This result of the investigation shows that on the one hand, interdiffusion of the two compounds of the uranium and the plutonium has taken place, i.e. that pure plutonium dioxide is no longer present but that, on the other hand, there is a lower limit in the series of possible compositions of the uranium and plutonium dioxide, above which also completely homogeneous solid solutions of the uranium and plutonium dioxide upon subsequent reprocessing with nitric aicd results in insoluble plutonium oxide. On the basis of the presently known results, it can be assumed that this lower limit is at a composition of about 50% uranium dioxide and about 50% plutonium dioxide. In this connection, it is worthy of note that it was found during the so-called post-irradiation investigation on mixed-oxide fuels irradiated in the reactor, that during the irradiation, demixing, i.e. separating out, of uranium and plutonium dioxide took place to a certain extent, with a preferred influence being observed in the temperature and stoichiometry gradient. This can have the result that, after irradiation, a higher concentration of plutonium is present in certain zones in the fuel rod than before the irradiation. Taking the investigation results mentioned above, into consideration, it can be expected in the case of mixed-oxide fuels with a plutonium content of more than 50% that fuels which were soluble prior to the irradiation exhibit insoluble components in certain zones after insertion into the reactor. SUMMARY OF THE INVENTION An object of the present invention is to provide a method for manufacturing mixed-oxide fuel pellets prepared by treating dry, solid uranium oxide and plutonium oxide which mixed-oxide fuel pellets are soluble in nitric acid and suitable for reprocessing without the aid of an additive such as hydrofluoric acid. Another object of the invention is to provide a method for producing a mixed-oxide pellet which has a solubility in nitric acid which is more than 99% with 95% certainty on the basis of measured values, referred to the plutonium content. A further object of the invention is to provide a method for the manufacture of mixed-oxide fuels with differently adjustable plutonium dioxide contents in a simple and efficient manner. With the foregoing and other objects in view, there is provided in accordance with the invention, a method for the manufacture of UO.sub.2 nuclear fuel pellets containing PuO.sub.2 in set amounts, which pellets are soluble in nitric acid, which includes (a) mixing uranium oxide powder having oxygen in stoichiometric excess of the dioxide, with plutnium dioxide powder in an amount of 15 to 50% plutonium dioxide by weight of the mixture of uranium oxide and plutonium dioxide, (b) milling the mixture of uranium oxide powder and plutonium dioxide powder and pressing the milled mixture to form pellets, (c) sintering the pellets in a reducing atmosphere in a furnace, communiting the sintered pellets to primary grain sizes of less than 2 .mu.m by milling, pressing the comminuted grains to form pellets, and comminuting the pellets to free-flowing granules, (d) mixing the free-flowing granules with uranium oxide granules in an amount to obtain a desired UO.sub.2 /PuO.sub.2 ratio in the resultant mixture, and (e) pressing the resultant mixture into pellets and sintering the pellets to form UO.sub.2 nuclear fuel pellets containing PuO.sub.2 soluble in nitric acid. In accordance with the invention, there is provided a method for the manufacture of UO.sub.2 nuclear fuel pellets containing PuO.sub.2 in set amounts suitable for fast nuclear reactors, which pellets are soluble in nitric acid, which includes (a) mixing uranium oxide powder having oxygen in stoichiometric excess of the dioxide, with plutonium dioxide powder in an amount of 15 to 50% by weight of the mixture of uranium oxide and plutonium dioxide to obtain a desired UO.sub.2 /PuO.sub.2 ratio in the resultant mixture suitable for producing UO.sub.2 nuclear fuel pellets containing PuO.sub.2 in set amounts for fast nuclear reactors, (b) milling the mixture of uranium oxide powder and plutonium dioxide powder and pressing the milled mixture to form pellets, (c) sintering the pellets in a reducing atmosphere in a furnace, comminuting the sintered pellets to primary grain sizes of less than 2 .mu.m by milling, pressing the comminuted grains to form pellets and comminuting the pellets to free-flowing granules, (d) and pressing the free-flowing granules into pellets and sintering the pellets to form UO.sub.2 nuclear fuel pellets containing PuO.sub.2 soluble in nitric acid and suitable for fast nuclear reactors. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a method for producing PuO.sub.2 /UO.sub.2 -nuclear fuel, it is nevertheless not intended to be limited to the details shown, since various modifications may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. |
046844961 | summary | CROSS REFERENCE TO RELATED APPLICATION Reference is hereby made to the following co-pending U.S. patent applications dealing with subject matter related to the present invention: 1. "Fuel Assembly Bottom Nozzle with Integral Debris Trap" by John F. Wilson, U.S. Ser. No. 672,041, filed Nov. 16, 1984. 2. "Wire Mesh Debris Trap for a Fuel Assembly" by William Bryan, U.S. Ser. No. 675,511, filed Dec. 7, 1984. BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactors and, more particularly, is concerned with a trap mounted in the bottom nozzle of a fuel assembly for capturing and retaining debris left in the reactor after assembly, repair and/or replacement operations and thereby preventing entry and lodging thereof in the fuel assembly where the debris can cause cladding perforations in the fuel rods and other damage to the fuel assembly. 2. Description of the Prior Art During manufacture and subsequent installation and repair of components comprising a nuclear reactor coolant circulation system, diligent effort is made to help assure removal of all debris from the reactor vessel and its associated systems which circulate coolant therethrough under various operating conditions. Although elaborate procedures are carried out to help assure debris removal, experience shows that in spite of the safeguards used to effect such removal, some chips and metal particles still remain hidden in the systems. In particular, fuel assembly damage due to debris trapped at the bottom grid has been noted in several reactors in recent years. The damage consists of fuel rod tube perforations caused by fretting of debris in contact with the exterior of the tube. The debris tends to be relatively thin rectangular pieces, as opposed to pieces which are spherical in shape. Specifically, most of the debris consists of metal turnings which were probably left in the primary system after steam generator repair or replacement. The debris lodges in the region of the lowermost grid within the spaces between its "egg-crate" shaped cell walls and the lower end portions of the fuel rod tubes. Almost all of the debris is deposited just above the four coolant flow openings in the lower core support plate. Several different approaches have been proposed and tried for carrying out removal of debris from nuclear reactors. Many of these approaches are discussed in U.S. Pat. No. 4,096,032 to Mayers et al. While all of the approaches described in this patent operate reasonably well and generally achieve their objectives under the range of operating conditions for which they were designed, a need still exists for a fresh approach to the problem of debris removal in nuclear reactors. The new approach must be compatiable with the existing structure and operation of the components of the reactor, be effective throughout the operating cycle of the reactor, and at least provide overall benefits which outweigh the costs it adds to the reactor. SUMMARY OF THE INVENTION The present invention provides a debris trap designed to satisfy the aforementioned needs. Underlying the present invention is the recognition of the mechanism by which debris is captured and retained in the bottom grid of the fuel assembly and application of that mechanism to the design of the trap structure. The entrapment mechanism by which small pieces of debris are trapped and retained by the lowermost grid centers around the interaction between the cell walls and compliant members (springs and dimples) of the grid and the fuel rod tube. The coolant flow seems to wedge the debris into the grid cell, and the compliance of the cell and debris holds it in place. The trap of the present invention seeks to simulate this mechanism but at a different location. Specifically, the trap is mounted within the bottom nozzle where coolant flow first enters the fuel assembly and employs a multiplicity of small cells with means for providing a lateral force on the debris to retain it in the cells. Thus, whenever coolant flow stops, the debris will not fall back into the system only to impact the trap structure again upon restart of coolant flow. By being mounted within the bottom nozzle, the trap structure and debris captured therein are removed from the reactor system with the fuel assembly. Also, the trap is designed to be backfittable on existing fuel assemblies. Accordingly, the present invention sets forth in a fuel assembly for a nuclear reactor including a plurality of nuclear fuel rods, at least one grid supporting the fuel rods in an organized array, an end nozzle disposed adjacent the grid and liquid coolant flowing through the end nozzle and into the fuel assembly, a trap for capturing and retaining debris carried by the flowing coolant to prevent entry of debris into the fuel assembly. The debris trap includes: (a) a structure disposed adjacent the end nozzle on an opposite side thereof from the grid, the structure forming a multiplicity of small cells each being open at opposite ends and defining a central channel for passage of coolant flow therethrough to the end nozzle; and (b) means defined in each of the cells for capturing and retaining within the structure any debris carried into the cells by the coolant flowing therethrough. More particularly, the debris trap structure includes interconnected wall portions forming each of the cells and defining the central channel thereof, with the capturing and retaining means being in the form of a spring-like finger attached to one of the wall portions and extending into the channel of the cell and downstream toward the end nozzle. Specifically, the structure is composed of a plurality of straps aligned with respect to each other in a crisscross interlocking arrangement and defining the wall portions of each cell in oppositely-disposed interconnected pairs thereof. One spring-like finger is punched out of each wall portion of each pair thereof, and the fingers on one pair of wall portions are disposed in the cell downstream of and in overlapping relation with the fingers on the other pair of wall portions. The structure is sized to fit within the end nozzle and includes means in the form of a pair of leaf springs for releasably locking it within the end nozzle. A plurality of tabs are mounted on the upstream end of the structure and project outwardly therefrom for grasping and holding pieces of debris generally stationary so as to prevent lateral circulation thereof along, and repeated impact thereof against, the debris trap structure. These and other advantages and attainments of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
051397323 | summary | FIELD OF THE INVENTION The invention relates to a process and a device for extracting a heating rod having deformations from a pressurizer casing of a pressurized-water nuclear reactor. BACKGROUND OF THE INVENTION Pressurized-water nuclear reactors have a primary circuit, in which the cooling water of the reactor is maintained at a high pressure of the order of 155 bars by means of a pressurizer arranged on one of the branches of the primary circuit. The pressurizer makes it possible to keep the pressure in the primary circuit between particular specific limits either by spraying, when the pressure tends to exceed the permissible upper limit, or by the electrical heating of the primary fluid, when the pressure tends to fall below the permissible lower value. These operations are conducted inside the pressurizer which comprises particularly a casing having a domed bottom, through which electrical heating rods are introduced inside the pressurizer. Passage sleeves are welded to the bottom of the pressurizer, each at a passage orifice. The heating rods are introduced into the sleeves and welded to their ends so as to ensure the mechanical stability and sealing of the connection. The heating rods are produced in the form of glove fingers containing the heating resistors and having an open end at the outer end of the passage sleeve, to ensure the connection and feed of the electrical resistors. Some heating rods can become defective during operation, and it is therefore necessary to check periodically that they are in a good operating condition. Should the presence of a defective rod be detected, its replacement is carried out so as to continue to ensure that the pressurizer operates satisfactorily. Conventionally, the replacement of a heating rod of a pressurizer is carried out by a sequence of operations including cross-cutting of the sleeve, extraction of defective rod, the cleaning and machining of the cut end of sleeve, the installation of a replacement rod, temporary fastening of this rod to the sleeve by means of manual welding spots, and permanent fastening of the rod by means of a continuous circular weld which ensures the junction between the rod and the end of the sleeve and is usually made by automatic welding. Various checking operations, to conducted between the different operations mentioned above, are all carried out from outside the pressurizer in a zone located underneath the lower domed bottom of its casing. The pressurizer casing of general cylindrical shape and closed by domed bottoms is arranged with its axis of symmetry in the vertical direction, and the heating rods are held in a vertical arrangement, i.e., parallel with the axis of the pressurizer casing, by means of spacer plates fixed to the inner wall of the casing. The heating rods, which are of limited length and which are arranged in the lower part of the pressurizer casing, are generally held by means of two spacer plates, namely, a lower spacer plate located above the bottom of the pressurizer and an upper spacer plate located above the lower spacer plate. Should a heating rod to be replaced have deformations, such as enlargements or bulges, the diameter of which is larger than the diameter of the through-holes for the heating rods in the spacer plates and larger than the bore of the sleeves, it is no longer possible to extract the rod from outside the casing of the pressurizer simply by an axial pull on the end of the heating rod. Such bulges of the metal casing of the heating rods usually occur in the region of the spacer plates and above the passage sleeves of the bottom of the pressurizer casing. To date, no process and device making it possible to extract deformed heating rods from the casing of a pressurizer has been known. SUMMARY OF THE INVENTION The object of the invention is, therefore, to provide a process for extracting a heating rod having deformations from a pressurizer casing of a pressurized-water nuclear reactor, which has an axis of symmetry and in which the heating rods are held in an axial direction by spacer plates and pass through a bottom of the casing inside sleeves, this process making it possible to extract the heating rod in order to ensure its replacement, regardless of the extent and arrangement of the deformations of the rod. To this end, the heating rod is cut inside the pressurizer casing in at least one zone by a remotely controlled, cutting operation, and at least one portion of the rod is extracted by way of an inspection port of the casing. The invention also relates to a device making it possible to extract a heating rod having deformations from the casing of a pressurizer. |
description | This disclosure relates to X-ray imaging systems, and, in particular, to X-ray imaging systems employing optic devices to produce X-ray beams having desired spectral shape and properties. Conventional laboratory sources produce a large cone of X rays, the majority of which typically are not utilized to analyze a sample in an X-ray system. X-ray optics may be used to redirect some of these unused X rays into useful directions. However, the efficiency of redirection decreases with increasing distance between the X-ray generation point inside the X-ray source and the collecting/redirecting optic(s). This decrease is typically due to the decrease in X-ray intensity with increasing distance between the X-ray generation point and the optics. Specifically, the X-ray intensity decreases as the square of the distance between the optics and X-ray generation point. Additional issues with respect to an X-ray source point and optics arrangement include: alignment of the optics and the X-ray generation point, and drifting of the X-ray beam generation point due to target heating or lack of dynamic electron beam control. The optimal position for disposing the optic to obtain maximum X-ray intensity would be at the X-ray generation point, however, with commercially available optics such as polycapillary optics or the singly- or doubly-curved diffractive optics such an arrangement is not feasible. Currently, polycapillary and multilayer diffractive optics are the only commercially available optics that collect a reasonable source solid angle and redirect the X rays into usable directions. The polycapillary channel size is too large to utilize small source spots, e.g. nanometer to micron-sized focal spots. Due to its large size, the polycapillary optics needs to be placed several hundred microns to centimeters away from the X-ray generation point. This large distance between the optics and the X-ray generation point reduces the intensity of the output X-ray beam that comes out of the optics. For both singly and/or doubly curved diffractive optics, the optics are typically placed at distances on the order of centimeters from the X-ray generation point to reduce the strain in the optics, which makes the optics difficult to manufacture and reduces the optic output beam quality. In addition to reducing the X-ray intensity, placing the optic at a large distance away from the X-ray generation point also reduces the robustness of the optic alignment with the X-ray generation point. Lastly, polycapillary optics are limited to transmitting X rays below 60 keV, which prevents them from being used in a number of non-destructive testing (NDT) imaging applications, e.g. computed tomography (CT) of electronic circuit boards. It would thus be desirable to dispose the optic device on the X-ray generation point so as to obtain optimal X-ray intensity, and to address alignment issues between the optic device and the target. In one embodiment, an integrated X-ray source is provided. The integrated X-ray source includes a target for emitting X-rays upon being struck by one or more excitation beams. Further, the integrated X-ray source includes one or more total internal reflection multilayer optic devices in physical contact with the target to transmit at least a portion of the X rays through total internal reflection to produce X-ray beams, where the optic device comprises an input face for receiving the X rays and an output face through which the X-ray beams exit the integrated X-ray source. In another embodiment, an integrated X-ray source is provided. The integrated X-ray source comprises one or more total internal reflection multilayer optic devices that are configured to receive electrons, wherein the optic devices comprise at least one low-index material layer and at least one high-index material layer, where a target material is interspersed in the low-index material layer. The target material is configured to produce X-rays upon being struck by one or more excitation beams, where the optic devices transmit at least a portion of the X rays through total internal reflection to produce X rays, wherein the optic devices comprise an input face for receiving the X rays and an output face through which the X rays exit the integrated X-ray source. In yet another embodiment, an X-ray imaging system is provided. The X-ray imaging system comprises a source of one or more excitation beams, and an integrated X-ray source. The integrated X-ray source includes a target for emitting X-rays upon being struck by the excitation beams from the electron source, and one or more total internal reflection multilayer optic devices in direct physical contact with the target to transmit at least a portion of the X rays through total internal reflection to produce one or more X-ray beams, wherein the optic devices comprise an input face for receiving the X-rays and an output face through which the X-rays exit the integrated X-ray source. Embodiments of the system relate to an integrated X-ray source having a target disposed in physical contact with one or more total internal reflection multilayer optic devices. The integrated X-ray source is configured to generate X rays and redirect the generated X rays via total internal reflection to produce X rays having desired beam shapes. The X-ray beams may include fan-shaped beams, beams with circularly symmetric cross-sections, or beams with elliptical cross-sections. Advantageously, the system enables greater X-ray flux output, and a more compact and robust design of the X-ray source. The optic device may include an input face and an output face. The input face may be defined as the face of the optic device that is closer to the incident excitation source, and the output face may be defined as the face through which the redirected X-ray beams exit the optic device of the integrated X-ray source. In one embodiment, the input face may be the face of the optic device through which the X-rays produced by the target enter the optic device. As used herein, the term “physical contact” encompasses presence of any additional material between the target and the optic device, where the material may be disposed between the target and the optic device to facilitate proper functioning of the integrated X-ray source. For example, the material may be disposed to facilitate coupling of the target and the optic device. In one example, an adhesive layer may be disposed between the target and the optic device. In one embodiment, there may not be any visible air gaps between the target and the optic device as seen by a naked eye. It is known that target focal spots are not completely static and can move dynamically, in some cases by tenths of a millimeter or more. Advantageously, the integrated X-ray source obviates the need for alignment of the target and the optic devices with respect to each other by providing an optic device that is integrated to the target. The creation of a single entity that acts both as an X-ray transmission target and has the capability of redirecting the generated X rays into useful directions results in increased X-ray beam intensity. Further, the size of the integrated X-ray source and the incident beam may be chosen such that the cross-section of the incident beam is larger than the cross-section of the target, the larger size of the incident beam ensures that the target is impacted by the incident excitation beams regardless of a shift in the position of the incident beam. Selecting suitable materials within the optic device may eliminate undesired energy levels from the output X-ray beam to produce X-ray beams having specific spectral properties. In one embodiment, the material selection within the optic device may reduce or eliminate high energies from the output X-ray beam, thereby allowing a bandpass of energies to exit the optic device. In one example, the energy widths of the output X-ray beam may be on the order of a few keV to a few tens of keV, if the input beam is for example a 100 kVp polychromatic spectrum. In one embodiment, the bandpass may be made sufficiently narrow to allow single photon energy to transmit through the optic device. In one embodiment, a k-edge filter may be used to reduce or eliminate low energies from the output X-ray beam. In this embodiment, the k-edge filter may be disposed either on the input face or the output face of the optic device. Alternatively, the k-edge filter may be disposed within the optic device. In embodiments where the k-edge filter is disposed on the input face, the filter may be disposed between the target and the optic device. In embodiments where the k-edge filter is disposed within the optic device, the k-edge filter may be disposed throughout the high-index material layers. Non-limiting examples of materials for the k-edge filter may include erbium. Combination of suitable filter and optic device materials may be used to allow single energy photons to be transmitted through the integrated X-ray source. In one example, the integrated X-ray source may allow 59.3 KeV of the tungsten K-alpha emission characteristic to be transmitted through the optic device. In some embodiments, the output X-ray beam may be a monochromatic X-ray beam. The monochromatic X-ray beam may be produced by employing k-edge filters with appropriate material selection of the optic device material to allow single energy to exit the optic device. In one example, a polychromatic X-ray beam may be produced by proper selection of material for the layers of the optic device. Further, the undesired energy ends may be minimized or eliminated using the total internal reflection within the optic device. Additionally, the optic could be dynamically cooled, thereby providing additional target cooling, again allowing the generation of a more intense beam than with a stand-alone transmission target. A minimum of three different materials are used in a graded multilayer stack to obtain increased total internal reflection over current practice by maximizing the difference in real refractive indices between successive layers, with the real refractive index decreasing in successive layers. In an embodiment that provides even greater total internal reflection, the ratio of the change in imaginary part of the refractive index to the change in real refractive index between successive layers is minimized by simultaneously minimizing the change in the imaginary part and maximizing the change in the real part of the refractive index between successive layers. The imaginary part of the refractive index is related to the mass-energy absorption coefficient of the material in which the X ray is traveling. Additionally, each successive layer has higher X-ray mass-energy absorption properties, while the real refractive index decreases monotonically from layer to layer. These criteria provide for optimal changes in real refractive index and X-ray absorption properties than in current reflective X-ray optics materials. Generally, the complex refractive index ‘n’ of a material at X-ray energies can be expressed as n=1−δ+iβ, where the term (1−δ) is the real part of the complex refractive index of the material and the parameter β is the imaginary part of the complex refractive index and is related to the mass-energy absorption coefficient in the material. At X-ray energies, the real part of the refractive index is very close to unity and is therefore usually expressed in terms of its decrement δ from unity, with δ typically on the order of 10−6 or smaller for energies above 60 keV. For improved reflectivity, in one embodiment, the ratio of the change in β to the change in δ between adjacent multilayer materials is generally minimized. For the purposes of this disclosure, a first layer is considered adjacent to a second layer when there are no other materials interposed between the first and second layers that have a real refractive index or a coefficient of absorption that are different from the respective real refractive indices or coefficients of absorption of the first and second layers. The graded multilayer optic may be adapted for use in redirecting an incident X-ray beam through total internal reflection as a reflected X-ray beam. The optic device may be configured to produce circularly symmetric beams, beams with elliptical cross-sections, or a stack of fan-shaped beams. The graded multilayer stack may comprise a plurality of multilayer zones. The graded multilayer optic device may be made by employing the techniques disclosed in the commonly assigned application titled “OPTIMIZING TOTAL INTERNAL REFLECTION MULTILAYER OPTICS THROUGH MATERIAL SELECTION” having application Ser. No. 12/469,121. In certain embodiments, the imaging system includes one or more graded multilayer optic devices in communication with the target to transmit at least a portion of the X rays through total internal reflection to produce one or more X-ray beams having desired shape and spectral properties. The graded multilayer optic devices include a first graded multilayer section for redirecting and transmitting X rays through total internal reflection. The first graded multilayer section includes a high-index layer of material having a first complex refractive index n1. The first complex refractive index n1 includes a real part Re(n1) of the first complex refractive index and an imaginary part β1 of the first complex refractive index. The real part Re(n1) of the first complex refractive index may also be represented as (1−δ1). The first graded multilayer section further includes a low-index layer of material having a second complex refractive index n2. The second complex refractive index includes a real part Re(n2) of the second complex refractive index and an imaginary part β2 of the second complex refractive index. The real part Re(n2) of the second complex refractive index may also be represented as (1−δ2). The first graded multilayer section also includes a grading zone disposed between the high-index layer of material and the low-index layer of material. The grading zone includes a grading layer having a third complex refractive index n3. The third complex refractive index n3 includes a real part Re(n3) of the third complex refractive index and an imaginary part β3 of the third complex refractive index. The real part Re(n3) of the third complex refractive index may also be represented as (1−δ3) such that Re(n1)>Re(n3)>Re(n2). As used herein, the term “imaginary part of the complex refractive index” corresponds to the mass-energy absorption coefficient. The target may be configured to emit X rays upon being struck by incident beams. The incident beams may include one or more of neutral particle beams, charged particle beams, or photon beams. Non-limiting examples of the target material may include tungsten, copper, silver, molybdenum, rhodium, or chromium. In some embodiments, the target may be made of a single material. In other embodiments, the target may include a plurality of materials such that at least one of the plurality of materials is an X-ray emitting material. In one embodiment, the X-ray emitting material may include one or more heat removing materials, or electrical discharge removing materials, or both. In one example, the target may be made of one or more radioactive materials. The target may either be a transmission target or a reflection target. In case of transmission target, the target may be disposed on the input face of the optic device. In case of transmission targets, the target may be disposed on the input face of the optic device such that the target and the input face of the optic device are in physical contact. In this case, the incoming excitation from the source strikes the target to produce X rays, the produced X rays are transmitted through the target layer to reach the optic device. The optic device then re-directs these X rays to produce X-ray beams having desired shape and spectral properties. In one embodiment, the target may be present in the form of a layered structure. The layered structure of the target may include a continuous layer or a patterned layer. Further, the layered structure may include a single layer or a plurality of layers. In the case of the plurality of layers, the different layers of the target may be made of the same or different materials. The layers of the target may be made of one or more materials. The materials of the target layers may be selected based on the X-ray energies desired. In the case of the plurality of layers, only some of the layers of the plurality of layers may be configured to emit X rays upon being impacted by the incident beams. For example, the plurality of layers may include an X-ray emitting layer disposed between layers that may at least partially remove heat and electrical charge from the X-ray emitting layer. In one example, at least one diamond layer, or at least one graphene layer may be disposed adjacent to the X-ray emitting layer. In one embodiment, the X-ray emitting layer may include tungsten, rhodium, molybdenum, rhodium-molybdenum alloy, copper, diamond, and alloys thereof. The dimensions of the target may be such that the target may efficiently stop the incident beam(s) impacting the target to produce X rays in the process, while minimally absorbing the generated X rays. The dimensions of the target may be large enough to prevent the target from suffering any structural damage that may be caused due to the impact of the incident beams. The dimensions of the target may be on the order of a few microns and may vary depending on the X-ray energies desired in the output X-ray beam and the selection of materials. In one embodiment, the target may be approximately perpendicular to the layers of the optic device. Further, the target may or may not have a uniform dimension (also referred to as “height”) in a direction perpendicular to the optic axis of the optic device. In one embodiment, the height of the target may be greater at its center than along the circumference/periphery. In an alternate embodiment, the target layer may include a stepped structure. In this embodiment, the target may include two or more steps. Each step may be in physical contact with a corresponding optic device. In other words, each of the steps may be associated with one or more optic device. The steps may be designed so as to produce multiple X-ray beams, with each of the X-ray beams having specific spectral and/or spatial properties. In one example, the plurality of X-ray beams may produce a focused X-ray beam, or a fan-shaped X-ray beam. Envisioning the layer of the target as a cuboid, the two opposite sides of the cuboid, one of which is coupled to the optic device, may or may not be parallel to each other. In one embodiment, the side away from the optic device may be skewed at a determined angle. In one example, the target may include a patterned structure. The patterned structure may be selected such that target material is disposed on those portions of the input face of the optic device that correspond to high-index layers of the optic device. In one example, the patterned structure may comprise a plurality of strips, where the strips containing the target material correspond to the portions of the input face having the high-index material or high-index and graded zone materials, and the gaps between the strips correspond to low-index materials. In other words, the pitch between the strips may be adjusted according to the height of the layers of the optic device. In one embodiment where the target includes a patterned structure, the low-index layer may be interspersed with the target material. For circularly symmetric optic devices, the strips comprising the target material may comprise concentric rings. In another embodiment, the low-index layer within the optic may be made from the target material. In one embodiment, a non-patterned target material is disposed on the front face of the optic, while the low index layers within the optic are made from the same or different target materials. This increases the intensity of the X-ray beam emitted by the optic, since the target material interior to the optic device can emit X rays in addition to the usual x rays emitted by the target material on the front optic face. In another embodiment, the target material for the patterned target is combined with low-index materials inside the optic being made from the same or different target material. The materials may be selected to produce a complex X-ray beam spectrum having specific desired spectral properties. In one example where the patterned target material is made from molybdenum, and the low-index materials are made from tungsten, the X-ray beam exiting the optic may have a well-separated (in energy) characteristic photon energies from each material. The well-separated energy spectrum is desirable, for example, in multi-energy imaging. The presence of high melting point and thermally conducting materials in the optic multilayer integrated with the target enables the optic device to act as a heat sink for the transmission target, permitting the target to be operated at higher flux densities than are normal for a transmission target, which is difficult to actively cool. Furthermore, the optic device may be placed in an actively cooled housing, providing further target cooling and further X-ray intensity increases. Non-limiting examples of thermally conducting materials may include diamond or diamond-like carbon (DLC). In one embodiment, one of the graded layers in the optic device may be made of thermally conducting material to provide thermal conductivity throughout the optic device. In an alternate embodiment, the thermally conducting layer may be made as part of the target disposed on the front face of the optic. In one example, such a thermally conducting layer may be made of diamond, since diamond is relatively transparent to high-energy (>60 keV) X rays. In case of reflection target, the target material may be present within the structure of the optic device. For example, the target material may be present in any of the layers of the optic device. For example, the target material may be present in the high-index material layer, or low-index material layer, or one or more layers of the grading zone. By way of example, the target material may be disposed in the low-index material layer of the optic device. In one embodiment, the low refractive index may be formed using the target material. In another embodiment, the low-index material layer may be interspersed or doped with the target material. In one example, the target may include a radioactive isotope. In this example, the low refractive index material layer of the optic device may be made of a radioactive isotope. In examples where the low-index material layer comprises the target material, the low-index material layer may include materials, such as but not limited to, tungsten, osmium and americium. In examples where the grading zone comprises the target material, the grading zone may include materials, such as but not limited to, gold, silver, molybdenum, cobalt, copper or chromium. In examples where the high-index material layer comprises the target material, the high-index material layer may include materials, such as but not limited to, magnesium, aluminum or silicon. In case of reflection targets, the optic device is pointed and spatially limited, the electric potential at the tip is higher than any surrounding support structure. The electron beam may be attracted to this high point compared to flat targets. This feature improves the positional stability of the X-ray generation point. In one embodiment, the integrated X-ray source may employ both transmission and reflection targets to produce monochromatic or polychromatic X-ray beams. In this embodiment, the target may be a combination of transmission and reflection targets. For example, the target may comprise a patterned structure with patterns disposed on portions of the input face of the optic device that corresponds to high-index layer, and the low-index layer may include target material disposed therein. In one example, the low-index layer may be made of low-index target material. One or more graded multilayer optic devices may be employed for redirecting and reshaping X rays generated by the target to produce monochromatic or polychromatic X-ray beams having desired spectral shape. The layers in the optic device may be shaped geometrically to collect a large solid angle of the generated X rays and redirect them via total internal reflection into determined directions for applications, such as but not limited to, computed tomography (CT) imaging, X-ray imaging, X-ray diffraction and X-ray fluorescence. In one example, the output X-ray beam may be a fan-shaped beam for application (medical, industrial and/or security) in one or more of CT imaging, X-ray imaging, tomosynthesis imaging, or X-ray diffraction imaging. In one example, the integrated X-ray source may be employed in high-resolution non-destructive testing (NDT) CT applications. In certain embodiments, all or a portion of the X rays from the source spot are physically shaped into a single, collimated, fan beam while intentionally altering the spectral distribution. For example, the spectrum may be altered to include only the very low (<30 KeV) and/or very high (>200 KeV) energy ends of the source spectrum. A plurality of optic devices may be stacked and in physical contact with the target to collect a majority (e.g., about 60 percent to about 90 percent) of the X rays from the target, and to produce a set of spatially shaped X ray beams. In one embodiment, the multilayer optic device may be circularly symmetric to generate a highly collimated beam in each spatial direction. In another embodiment, a stack of parallel fan beams is produced with the parallel direction perpendicular to the plane of the fan. In one embodiment, the graded multilayer optic devices may include pairs of stacked graded multilayer optic devices. In one example, one half of a pair may be positioned to be a mirror image of the other half of the pair. The graded multilayer optic sections stacked upon each other may have an exterior surface sloping between an input and an output face. In certain embodiments, each layer at the optic input (side closest to the source) may be curved at the same or different radius of curvature enabling the combined layers in the optic device to capture a large source solid angle (see e.g., FIG. 3). The number of multilayer zones comprising the multilayer material stack is not limited in any way but is rather a function of the particular application for which the multilayer material stack is configured. The multilayer material stack may comprise tens or thousands of multilayer sections. For example, in the case of high-resolution industrial CT where the resolution is on the order of micrometers, the number of multilayers in the stack maybe less than ten layers. In other types of CT, where large optic collection angles are desired, the number of layers may be in the thousands. In addition to a high-index layer, a low-index layer, and a grading zone with one or more grading layers disposed between the high-index layer and the low-index layer, the multilayer optic device may also comprise an X-ray opaque cladding layer at the outermost surface of the optic device to prevent the emission of X-ray radiation from the interior of the optic device through the edges of the non-emitting face of the device. The X-ray opaque cladding layer may be disposed on the optic device such that X-rays enter the optic device through the input face and exit the optic device substantially through the optic output face. Typically, high refractive index materials transmit X rays with minimal losses, whereas, low refractive index materials substantially block X-ray transmission. In the case of the interspersed target, the target material may be disposed in one or more low-refractive index material layers of the optic device. The amount of the target material present in the low refractive index material layer of the optic device may be decided based upon the desired output X-ray flux, or intensity. It may be desired to prevent the incident source excitation (photon or electron) from impinging on some of the high refractive index material layers of the optic device. In one embodiment, a blocking layer may be disposed on the input face of the optic device to selectively block the source excitation from reaching specified portions of the optic device. In one example, the blocking layer may be selectively disposed on portions of the input face of the optic device that correspond to the high refractive index material layer. The incident excitation thus impacts only the low refractive index material layer. The target material disposed in the low refractive index material layer enables production of X rays upon interaction with the incident excitation. When the produced X rays encounter an interface between a high and low refractive index material, the X rays may be reflected via total internal reflection back into the high refractive index material with high efficiency, if the X rays are traveling from the low to high refractive index material. The value of the critical angle for total internal reflection depends on the materials and the incident X-ray energy. The use of the graded multilayer optic device enables X rays of desired energies to be reflected via total internal reflection with high efficiency. Shaping the layers with the appropriate curvature and fabricating them with the appropriate heights may produce output beams having desired properties. The thickness of the incident X-ray beam may be smaller or greater than a height of one multilayer section. When the thickness of the incident X-ray beam is greater than the height of one multilayer optic device, different parts of the incident X-ray beam may pass through and be totally internally reflected by some or all of the multilayer sections within the optic device, and emerge from the multilayer sections as corresponding parts of the reflected photon beam. Alternatively, when the thickness of the incident photon beam is smaller than the height of one optic device, the device may produce smaller flux gains but can provide useful redirection capabilities. The optic device provides an advantage in terms of spatial scale and flexibility of the integrated X-ray source. Due to the nature of the micro-fabricated, layered structure, the optic devices may be very small. In one example, a cross sectional size of the devices may be as small as tens of micrometers. The samples to be imaged are typically disposed at a distance of about 1 meter or more from the X-ray integrated source, hence, the output X-ray beam from the source needs to diverge from about 1 mm to about several centimeters or more. Advantageously, the integrated X-ray source is configured to reduce the source power needed to produce the same sized beams as produced by conventional sources at the sample. The integrated X-ray source may be enclosed within a housing having an X-ray transparent window. In the case of transmission targets, the integrated X-ray source may be mounted within or exterior to the housing. In one embodiment, the target may be optically coupled to the window, either interior or exterior to the housing. In one example, the integrated X-ray source may form part of the window. In the case of reflection targets, the optic device of the integrated X-ray source may be located almost exclusively internal to the source vacuum, with its output face towards the X-ray window through which the X rays exit the source housing. In one example, the optic device may serve as the X-ray window through which the X rays exit the source. Advantageously, replacing the conventional X-ray window with the optic device simplifies the source design and provides significantly greater X-ray flux output than without the optic device. FIG. 1 is a diagrammatical cross-sectional illustration of an exemplary embodiment of an integrated X-ray source 10. The integrated X-ray source includes a target 12 in physical contact with a total internal reflection multilayer optic device 14. The target 12 may be a layered structure having one or more layers. The number and material of the layers of the target 12 may be governed by the incident excitation (electrons or photons, e.g., X rays or gamma rays) and the desired output X-ray beams 16. The target 12 may be a continuous layer or a patterned layer, the structure of the target 12 may depend on the material used in the plurality of layers of the optic device. The optic device 14 includes high-index layers 18 and low-index layers 20 that are alternatingly disposed. Reflecting interfaces are formed between each pair of high-index layers 18 and low-index layers 20. The height (h) and length (l) of the target may be such that the target may efficiently stop the incident beam(s) impacting the target to produce X rays in the process, while minimally absorbing the generated X rays. Incident excitation beams 24 may be provided by an excitation source (not shown). The incident excitation beams 24 may include particle beams—for example electrons—or photon beams—such as X rays or gamma rays. The incident excitation beams 24 are directed to the X-ray integrated source 10 to impact on an input face 26 of the optic device 14. X-ray beamlets 27 are generated as a result of the incident excitation beams 24 striking on the target layer 12. The X-ray beamlets 27 are transmitted through the target 12 into the optic device 14, where the X-ray beamlets 27 follow transmission paths defined within the optic device 14. Using the transmission paths of the optic device 14, the X-ray beamlets 27 are guided towards the output face 28 of the optic device 14. The output X-ray beam 16 may be a parallel beam, or any other beam shape depending on the curvature of the optic device. Although the output X-ray beam 16 is shown in the illustration as separate parallel X-ray beamlets 27, it should be understood that the X-ray beam 16 is physically a continuous beam distributed over a specified solid angle of emission, and that the representation of the X-ray beam 16 as discrete beamlets is made only to facilitate the presentation of the various exemplary embodiments herein. The height (H) of the optic device 14 may be a sum of individual heights of the various layers of the optic device 14. In one embodiment, the height of an individual high refractive index layer 18 may be in a range from about 3 nm to about 50 nm. The height of an individual low refractive index layer 20 may be in a range from about 1 nm to about 10 nm. In one example, the height of the high refractive index layers 18 may be about 150 nm, and the height of the low-index layers 20 may be about 3 nm. The height (H) and length (L) of the optic device 14 may be decided based on the transmission path required for X-ray beamlets 27 to provide a desired beam shape at the output face 28 of the optic device 14. For ease of illustration, only a few layers have been drawn with reference to multilayer optic 10. However, it should be appreciated that any number of layers, including into the hundreds, thousands, or millions of layers, can be fabricated to utilize total internal reflection to form the various types of photon beams listed previously. In the illustrated embodiment, the central axis of the optic device 14 is coincident with a central axis 29 of the target 12. In embodiments where the optic device 14 is symmetric about the central axis, the resultant beam may be a concentric beam uniform about the central axis 29. The beam shaped may be circularly symmetric about the central axis 29. However, it should be noted that other beam arrangements such as non-concentric or non-circular asymmetric beam shapes may also be produced using the integrated X-ray source of the system. Referring now to FIG. 2, an integrated source comprising a target 62 and a multilayer material stack 30 is illustrated. The multilayer material stack 30 includes an input face 64 and an output face 66. The multilayer material stack 30 further comprises first and second multilayer zones 32-1 and 32-2, each multilayer zone comprises multiple layers of materials, each layer of material having a unique real refractive index n, an absorption coefficient β, and a height h. In the example provided, the multilayer zones 32-1 and 32-2 each include: (i) the high-index layer 34 with a real refractive index n1, an absorption coefficient β1, and a height h1; (ii) the grading zone 36 with a plurality of grading layers, here represented by grading layers 36-1 through 36-3, and (iii) the low-index layer 38 with a real refractive index n2, an absorption coefficient β2, and a height h2, disposed on the grading zone 36. The material forming the first grading layer 36-1 has a real refractive index n3, an absorption coefficient β3, and a height h3, disposed on the high-index layer 34. The material forming the second grading layer 36-2 has a real refractive index n4, an absorption coefficient β4, and a height h4, disposed on the first grading layer 36-1, and the material forming the third grading layer 36-3 has a real refractive index n5, an absorption coefficient β5, and a height h5, disposed on the second grading layer 36-2. The heights of the high-index layer 34 and the low-index layer 38 may typically be on the order of nanometers to microns depending on the desired output beam divergence, and the heights of the grading layers 36-1 through 36-3 may typically be on the order of nanometers to microns also. It should be understood that the number of multilayer zones comprising the multilayer material stack 30 are not limited in any way but is rather a function of the particular application for which the multilayer material stack 30 is configured. The multilayer material stack 30 may comprise hundreds or thousands of multilayer zones. Each multilayer zone 32-1 through 32-N includes a high-index layer 34, a low-index layer 38, and a grading zone 36 with one or more grading layers disposed between the high-index layer 34 and the low-index layer 38. The material layers making up each multilayer zone are selected and arranged in accordance with methods described herein. The multilayer material stack 30 may also comprise a photon-opaque cladding layer 44 at an outer surface of the multilayer material stack 30 to prevent the emission of photon radiation from the 1St multilayer zone 32-1. X-ray beam 40 is produced as a result of incident excitation beams 60 striking on the transmission target 62. The incident excitation beams 60 is comprised of excitation beams (e.g. electron or gamma-rays) beams striking the target 62 for generating X-ray beams 40. A first part of the X-ray beam 40 undergoes total internal reflection at a first interface 50, formed between the high-index layer 34 and the first grading layer 36-1, and emerges from the multilayer material stack 30 as a first beamlet 42-1. Most of the incident photon beam 40 is reflected by the first and second multilayer zones, 32-1 and 32-2, but very weak beamlets 42-6 and 42-7 pass into the next multilayer zones (not shown). Note also that the illustration is not drawn to scale, and that the material layer heights and the angles of incidence and reflection for the incident photon beam 40 are exaggerated for clarity of illustration. A second part of the incident photon beam 40 reflects at a second interface 52 and emerges from the multilayer material stack 30 as a second beamlet 42-2. In an exemplary embodiment, the second beamlet 42-2 has a different intensity from that of the first reflected beamlet 42-1, and is usually of much lower intensity. For example, an even lower intensity, third part of the incident photon beam 40 may reflect at a third interface 54 and emerge from the multilayer material stack 30 as a possibly even lower intensity third beamlet 42-3. Similarly, a yet lower intensity, fourth part of the incident photon beam 40 may reflect at a fourth interface 56 and emerge as a still lower intensity fourth beamlet 42-4. And, in the illustration provided, a still lower intensity, fifth part of the incident photon beam 40 may reflect at a fifth interface 58 and emerge as a yet lower intensity beamlet 42-5, leaving a negligible portion of the incident beam 40 to pass into the next multilayer zone (not shown) as the spurious, very low intensity, beamlet 42-6. FIG. 3 illustrates an isometric diagrammatical representation of the multilayer zone 32-2 of the optic device 30 of FIG. 2. In the illustrated embodiment, the high-index core 34 may comprise a rod-like structure. The cross-section of the core may or may not be circular. For example, in alternate embodiments, the core may have a hexagonal, rectangular, square, or any other geometric shape cross-sectional area. The cross-sectional shape of the high-index core 34 may be determined by the particular geometry of a high-index fiber material, such as beryllium or boron used for fabrication of the high-index core 34, without effecting functionality of the optic device 30. It should be noted that the concentric multilayer zone 36-1 through 36-3 is not restricted to three layers, and may have hundreds or thousands or millions of concentric multilayer zones. A portion of the outer low-index layer 38 comprises a convex surface curved toward the longitudinal axis of the optic device. It should be noted that in place of the convex surface, the optic device 30 may comprise a saddle surface (not shown) close to the output face 66 of the optic device 30. In one embodiment, the saddle surface may result in the input face of the optic device having a relatively larger cross-sectional area compared to the output face. The presence of saddle surface in the optic device may allow for conversion of a convergent input beam into a substantially collimated output beam. Further, the input face may be planar or curved in concave or convex or complex curved shapes. Cylindrical grading layers 36-1, 36-2 and 36-3 physically enclose the high-index core layer 34, and an outer low-index layer 38 encloses the grading layers 36-1, 36-2 and 36-3. FIG. 4 comprises an integrated X-ray source employing a reflection target and a multilayer optic device 70. The reflection target is present within the optic device 70. The optic device comprises an input face 71 and an output face 73. The optic device comprises a core 72 made of high-index material, grading zones 74 having grading layers and low-index layers 78. The low-index layers 78 may be doped or interspersed with the target material. Alternatively, the low-index layer may be made of target material. Non-limiting examples of low-index materials may include osmium, tungsten. In one embodiment, the different low-index layers may comprise different target materials. When such an integrated X-ray source is bombarded with incident excitation beams, two different spectra of X-rays corresponding to the two different target materials may be generated. In another embodiment, each of the low-index layers 78 may comprise two or more different target materials having low refractive index. In this embodiment, portions of the input face 71 corresponding to the high-index core 72 and the grading zones 74 may comprise a blocking material, such as but not limited to an absorber or a reflector material disposed thereon. The blocking material minimizes or prevents any damage to the material of the optic device 70 which may be otherwise caused due to impact of the striking incident excitation beams. FIG. 5 illustrates an arrangement for producing polychromatic X-ray spectra using integrated X-ray source 100. The integrated X-ray source 100 includes a plurality of optic devices 110. The structure of the optic devices 110 may be the same or different. The transmission target 112 comprises steps 114. Each of the steps may be designed to enable generation of X-rays of a particular energy level upon being struck by electron beam 120. The corresponding optic devices 110 may be configured to redirect and reshape X rays of a particular energy. The low-index material layers of the optic devices 110 may or may not include target material. The length of steps 114 may be less than a micron or as large as tens of millimeters or more. Large size of the steps 114 may increase the flux incident on the input face of optic devices 110. By increasing the length of steps 114 and increasing the incident angle (e.g., approaching 90 degrees) of the electron beam 120 on the steps 114, flux incident on the input surface of the optic device may be increased. Increased incident flux enables higher intensity X-ray beams 118 at the output face of the optic devices 110. Each of the steps 114 may be maintained at a particular potential. In one embodiment, greater energy separation between the X rays generated in the different optic devices 110 may be achieved by maintaining the steps 114 at different accelerating potentials and taking sequential images using the detector 116 with X rays 118 emitted by each accelerating potential/optic combination. Advantageously, spectral shaping facilitates optimizing the effectiveness of a multitude of X-ray inspection and scanning procedures otherwise required in CT, X-ray radiographic, or X-ray diffraction applications. Although discussed in terms of considering differences of images at multiple energies, as will be appreciated by one skilled in the art, standard projection-based and image-based energy sensitive decomposition methods may be utilized to characterize the effective atomic number of the imaged objects. The applications in which this system is anticipated to be used are non-destructive testing ones that currently suffer from insufficient x-ray flux, making data collection times impractical for industrial in-line testing. These NDT applications are anticipated to be high-resolution CT, X-ray radiography, x-ray diffraction, and x-ray fluorescence. In the specific case of NDT x-ray diffraction (as opposed to Security-related XRD), a circularly symmetric beam is used that is either monochromatic (usually for stress/strain measurements or powder samples) or polychromatic (for single crystal Laue measurements or energy-sensitive diffraction). In the monochromatic situation, typically a crystal is used to monochromate a polychromatic beam and in the process decreases the x-ray flux on the sample by between one and three orders of magnitude depending on the type of optics used. This low flux makes the diffraction measurement time too long for in-line inspection of parts, something that industry desires. In the Laue case, the beam must be very parallel and is accomplished through the use of a collimator that simply blocks the diverging x rays going in the wrong direction. Thus, the X-ray flux intensity is reduced by several orders of magnitude when imaging the sample, making the measurement times unreasonable for in-line parts inspection. This written description uses examples to disclose the invention and also to enable any person skilled in the art to practice the invention, including making and using any devices or systems and performing any incorporated methods. The patentable scope of the invention is defined by the claims, and may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if they have structural elements that do not differ from the literal language of the claims, or if they include equivalent structural elements with insubstantial differences from the literal languages of the claims. |
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06233302& | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS One conventional type of nuclear reactor in which the MOX fuel can be utilized is the pressurized water reactor (PWR). This type of reactor typically combusts a uranium oxide (UO.sub.2) fuel to produce steam. These NSSS (nuclear steam supply systems) traditionally include two-loop arrangement with two steam generators, two hot legs, and four cold legs each with a reactor coolant pump. One specific example of a PWR in which the embodiments of the invention can be implemented is ABB Combustion Engineering's System 80.RTM. reactor which loads 241 fuel assemblies. Each assembly, as seen in FIG. 1, is mechanically identical to the others and consists of a 16.times.16 fuel rod array 20 with five large structural guide tubes 21 that each occupy 2.times.2 fuel lattice locations. The four outer guide tubes are for control element assembly (CEA) fingers, while the center guide tube is used for in-core instrumentation. The in-core instruments are bottom-entry, and therefore do not interfere with the upper internal design for CEA guidance. Each fuel assembly contains 236 fuel rods 22. As seen in FIG. 2, the CEA's have either 4 or 12 element arrangements. The 12 element CEA has the unique characteristic of inserting into five adjacent fuel assemblies, as shown in FIG. 3. This characteristic is made possible by the unique upper guide structure design. of the reactor internals, which provide continuous guidance for each individual CEA element into the fuel assembly guide tube. This upper guide structure, shown in FIG. 4, is a rugged, all-welded structure, and protects each individual CEA element from flow forces and dynamic loads. In this UO.sub.2 core design, burnable absorber pins which contain erbia (Er.sub.2 O.sub.3) admixed with enriched UO.sub.2 are used in the fuel assemblies. These burnable fuel rods are located in predetermined locations to provide reactivity hold down and control power peaking. Reference will now be made in detail to the preferred embodiments of the invention, examples of which are illustrated in the accompanying drawings. FIGS. 5A to 14A show ten different 16.times.16 fuel assembly designs containing MOX fuel which were developed for use in the equilibrium cycle core designs of the above mentioned type of nuclear reactor. According to the invention, it is possible to select a number (e.g. 241) of one or more of the these fuel assembly designs and to compile the core in a manner which will enable a particular set of combustion characteristics, such as produced by using only conventional uranium fuel, to be replicated. Burnable absorber rods containing erbia are incorporated into these MOX fuel assembly designs to provide reactivity hold down and control power peaking. These are fuel-bearing burnable absorbers, but do not contain MOX in accordance with the above mentioned ground rules/restraints which have been imposed. Instead, the burnable absorber rods employed in these MOX assemblies are, in the disclosed embodiments, an admixture of erbia and enriched UO.sub.2, and are functionally identical to the burnable absorber pins described earlier in the discussion of the traditional UO.sub.2 core design. The fuel assembly designs in FIGS. 5 to 14 are differentiated by the number of MOX fuel rods and the number of urania-erbia (UO.sub.2 --Er.sub.2 O.sub.3) rods within each assembly as well as by the specific arrangement of these rods. In FIGS. 5 to 14, "M" represents a MOX fuel rod and "E" represents an urania-erbia fuel rod. The number of urania-erbia rods in the fuel assembly designs in the arrangements shown in FIGS. 5 to 14 ranges from 24 to 88. Within each fuel assembly design, the locations of the burnable absorber (urania-erbia) rods and the MOX fuel rods are fixed. Both the UO.sub.2 enrichment in the urania-erbia rods and the plutonium enrichment (wt % of Pu-239) in the MOX fuel rods can be varied during the core design process. Typically, there are 5 to 8 different plutonium enrichments in the MOX fuel rods within any given fuel assembly. For the urania-erbia rods, the UO.sub.2 enrichment is the same in all of the rods within a particular fuel assembly. Each of the fuel assembly designs in FIGS. 5A to 14A were developed on an octant basis and are octant-symmetric. Each of FIGS. 5B to 14B and 5C to 14C depict, for an assembly octant, the specific Pu-239 enrichment of each MOX fuel rod and the resulting normalized intra-assembly power distribution. Since the enrichment of the burnable absorber rods is fixed within any one of these fuel assembly designs, the respective octant maps in FIGS. 5B to 14B and 5C to 14C identify them within each assembly with the letter "E". Actually, two such octant maps are depicted for each. assembly design, representing data for a low enrichment case in FIGS. 5B to 14B and a high enrichment case in FIGS. 5C to 14C, respectively. Between these two cases, each fuel pin's enrichment differs by exactly 1.0 wt. % For the MOX pins, the Pu-239 enrichment is as shown. For the erbia pins, a fixed UO.sub.2 enrichment of 4.0 wt. % is selected for the low enrichment case and a fixed UO.sub.2 enrichment of 5.0 wt. % is selected for the high enrichment case. Each MOX assembly is designed to provide optimal performance over this range of enrichments represented by the low enrichment case and the high enrichment case. Detailed neutronics, generated for both cases, indicates that the neutronics behavior is characterized as a function of fuel enrichment. This design approach makes it possible to consider the effects of varying assembly enrichments during an equilibrium cycle core design phase without the need of re-generating any additional assembly data. By using different fuel rod enrichments within each MOX fuel assembly as described herein and as shown in the corresponding figures, it is possible to both optimize the intra-assembly power peaking, which enhances the performance of the fuel assemblies during operation, and to maximize the throughput of weapons-grade plutonium in each core. The burner absorber rod characteristics for the MOX assembly designs are also arranged to optimize the intra-assembly power peaking and have the secondary benefit of enhancing the throughput of weapons-grade plutonium in each core. FIG. 5A shows the MOX fuel assembly design of a first embodiment of the instant invention having a 16.times.16 fuel rod array including 24 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 5B shows a low enrichment octant map and FIG. 5C shows a high enrichment octant map of this embodiment. Each of these maps depicts one octant of the 236 rod arrangement shown in FIG. 5A. As will be noted, in the case of the low enrichment, while most of the MOX rods have a Pu-239 enrichment of 4.8wt %, a number of the rods, which are in proximity of the guide tubes 21, have lower values which are as low as 3.3wt %. The corresponding MOX rods according to the high enrichment schedule are, as mentioned above, 1% richer. FIG. 6A shows the MOX fuel assembly design according to a second embodiment of the instant invention and which has a 16.times.16 fuel rod array including 32 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 6B shows a low enrichment octant map and FIG. 6C shows a high enrichment octant map of this embodiment. FIG. 7A shows the MOX fuel assembly design of a third embodiment of the instant invention having a 16.times.16 fuel rod array including. 40 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 7B shows a low enrichment octant map and FIG. 7C shows a high enrichment octant map of this embodiment. FIG. 8A shows the MOX fuel assembly design of a fourth embodiment of the instant invention having a 16.times.16 fuel rod array including 48 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 8B shows a low enrichment octant map and FIG. 8C shows a high enrichment octant map of this embodiment. FIG. 9A shows the MOX fuel assembly design of a fifth embodiment of the instant invention having a 16.times.16 fuel rod array including 56 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 9B shows a low enrichment octant map and FIG. 9C shows a high enrichment octant map of this embodiment. FIG. 10A shows the MOX fuel assembly design of a sixth embodiment of the instant invention having a 16.times.16 fuel rod array including 60 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 10B shows a low enrichment octant map and FIG. 10C shows a high enrichment octant map of this embodiment. FIG. 11A shows the MOX fuel assembly design of a seventh embodiment of the instant invention having a 16.times.16 fuel rod array including 64 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 11B shows a low enrichment octant map and FIG. 11C shows a high enrichment octant map of this embodiment. FIG. 12A shows the MOX fuel assembly design of an eighth embodiment of the instant invention having a 16.times.16 fuel rod array including 72 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 12B shows a low enrichment octant map and FIG. 12C shows a high enrichment octant map of this embodiment. FIG. 13A shows the MOX fuel assembly design of a ninth embodiment of the instant invention having a 16.times.16 fuel rod array including 80 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 13B shows a low enrichment octant map and FIG. 13C shows a high enrichment octant map of this embodiment. FIG. 14A shows the MOX fuel assembly design of a tenth embodiment of the instant invention having a 16.times.16 fuel rod array including 88 erbium (UO.sub.2 --Er.sub.2 O.sub.3) rods. FIG. 14B shows a low enrichment octant map and FIG. 14C shows a high enrichment octant map of this embodiment. In accordance with the invention, equilibrium cycle core designs using MOX fuel can be developed using a subset consisting of any combination (e.g. up to three) of the ten fuel assembly designs shown in FIGS. 5A to 14A. FIGS. 15 and 16 show examples of two different equilibrium cycle core loading patterns having a feed batch size of 81 fuel assemblies (i.e. 81 new fuel assemblies). FIG. 17 shows an equilibrium cycle core loading pattern having a feed batch size of 88 fuel assemblies. In FIGS. 15 to 17, "X" represents a fresh assembly, "Y" represents a once-burned assembly and "Z" represents a twice-burned assembly, "O" represents an assembly sub-type with 24 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods, "3" represents an assembly sub-type with 48 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods, "4" represents an assembly sub-type with 56 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods, "5" represents an assembly sub-type with 60 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods and "7" represents an assembly sub-type with 72 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods. FIG. 15 shows a feed batch having 25 assemblies with 24 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods and 56 assemblies-with 56 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods. The average enrichment of MOX fuel rods in this feed batch is 5.16 wt % Pu-239. FIG. 16 shows a second feed batch arrangement having 17 assemblies with 24 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods, 36 assemblies with 56 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods and 28 assemblies with 72 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods. The average enrichment of MOX fuel rods in this feed batch is 5.01 wt % Pu-239. FIG. 17 shows a third feed batch arrangement having 64 assemblies with 48 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods, 12 assemblies with 56 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods and 12 assemblies with 60 UO.sub.2 --Er.sub.2 O.sub.3 fuel rods. The average enrichment of MOX fuel rods in this feed batch is 4.67 wt % Pu-239. A summary of some important design parameters for the equilibrium cycle for a MOX core design and a typical UO.sub.2 core design is shown in Table 1 set forth at the end of this disclosure. These equilibrium cycle core designs using MOX fuel were evaluated to assess their performance characteristics relative to a typical UO.sub.2 equilibrium cycle core design. As will be appreciated, the invention enabled the MOX core to perform in a manner which closely corresponds to the power level--average coolant temperature series of parameters produced with a UO.sub.2 core. Table 2, which is also set forth at the end of this disclosure, shows a comparison of some important core performance characteristics for the 88 feed batch assembly MOX core design shown in FIG. 17 and a typical 18-month cycle UO.sub.2 core design. The core average burn-up for the MOX-based 18-month cycle core design (17,000 MWd/MTHM) is consistent with that for a similar UO.sub.2 -based cycle (17,500 MWd/MTHM). The maximum fuel rod burn-up is within the licensed limit of 60,000 MWD/MT. The discharge burn-up (46,000 MWd/MTHM) is consistent with discharge burn-ups (45,000 MWd/MTHM) for comparable UO.sub.2 -based fuel cycles. The hot full power (HFP) all-rods-out (ARO) BOC critical boron concentration (CBC) for the MOX-based core. design is 1990 ppm, compared to 1250 ppm for a UO.sub.2 -based core design. Although larger than the value for a typical UO.sub.2 core, the HFP BOC CBC for the MOX core is less than the maximum allowable value of 2000 ppm necessary to remain within the existing analysis envelope for existing plants. The power distributions for the MOX-based core design are similar to those for a comparable UO.sub.2 -based core design. The maximum expected peaking factors for the MOX core are slightly higher than those for the UO.sub.2 core, but within the allowable limits (less than or equal to 1.72 for Fr, 2.00 for Fz, for HFP ARO conditions) necessary to remain within the existing analysis envelope. The MOX-based equilibrium cycle core designs developed in the instant invention achieve a throughput of approximately 1.5 MT (Metric Tons) of weapons-grade plutonium per 18 month cycle. As a result, the disposal of 50 MT of weapons-grade plutonium could be accomplished in three System 80.RTM. reactors in approximately 17 years of plant operation. This includes the transition from a conventional, low enrichment UO.sub.2 core to a MOX core. Numerous modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that within the scope of the appended claims, the invention may be practiced otherwise than as specifically described herein. For example, as will be self-evident from the above disclosure, if a fuel qualification for MOX fuel wherein the use of erbia in the MOX rods was permitted, a substantial increase in the throughput of plutonium would be enabled with an attendant reduction in the time needed to dispose of any given quantity of weapons grade plutonium. TABLE 1 Core Design Parameters UO.sub.2 Core MOX Core Power Level 43876 Mwth 3876 Mwth Nominal Cycle Length 18 Months 18 Months .about.460 EFPD 463.5 EFPD 17,500 GWd/MTHM 17,000 GWd/MTHM Fuel Assemblies 241 241 Fuel Assembly Configuration 16 .times. 16 16 .times. 16 Fuel Rod Locations/Assembly 236 236 Active Core Height (inches) 150 150 Fuel Loading (MTHM) 102.3 102.3 Fuel Type Enriched U-235 Enriched WG Pu.sub.2 O.sub.3 in Tails UO.sub.2, Burnable Absorber Type Er.sub.2 O.sub.3 in Enriched UO.sub.2 Er.sub.2 O.sub.3 in Enriched UO.sub.2 Fuel Management 3-batch, mixed central zone 3-batch, mixed central zone Erbia Loading (integral) <2.5 wt % (integral) <2.5 wt % Feed Batch Size Assemblies 72-104 81-88 Feed Fuel Enrichment <4.5 wt % U-235 .about.4.5-5.0 wt % Pu-239.sup.(1) Soluble Burnable Absorber Natural B.sub.10 Natural B.sub.10 Control Element Assemblies Standard Configuration 76 Full-Length, Full Strength 76 Full-Length, Full Strength 13 Part-Length, Part Strength 13 Part-Length, Part Strength or Enhanced Configuration 89 Full-Length, Full Strength Average Heat Generation Rate 5.45 KW/FT 5.45 KW/FT Average Coolant Temperature 585 .degree. F. 585 .degree. F. .sup.(1) Average for MOX pins only. TABLE 2 Core Performance Characteristics System 80 .RTM. Equilibrium Cycle Core Design UO.sub.2 Core MOX Core BOC EOC BOC EOC Burnup Data, MWd/MTHM Core Average 13,700 31,200 17,000 34,000 Maximum Fuel Rod -- 51,600 -- 57,400 Discharge Batch Average -- 45,000 -- 46,400 Critical Boron Data, PPM HFP, ARO 1250 1 1990 90 Inverse Boron Worth, PPM/% .DELTA..rho. Hot Full Power -130 -107 -227 -169 Maximum Peaking Factors Fr (HFP, ARO) 1.51 1.64 Fq (HFP, ARO) 1.83 1.96 Moderator Temperature Coefficent (MTC), 10.sup.-4 .DELTA..rho./.degree. F. Hot Zero Power +0.17 -- -1.62 -- Hot Full Power -0.72 -2.89 -1.80 -3.91 Standard CEA Standard CEA Enhanced CEA Configuration Configuration Configuration CEA Worths, % .DELTA..rho..sup.(1) Total Net Total Net Total Net BOC, HZP 12.6 10.0 11.0 8.7 11.9 9.9 EOC, HZP 15.1 11.3 13.3 10.2 14.4 11.3 EOC, Cold (68.degree. F.) 11.1 7.5 10.2 6.7 11.1 7.7 .sup.(1) The CEA worths are raw values, with no biases or uncertainties, for comparison purposes only. |
summary | ||
summary | ||
06014422& | abstract | The present invention provides combining the advantages of hybrid resist with the unique properties of x-ray lithography to form high tolerance devices with x-ray pitch and to provide a means for varying the space width and fine tuning to account for process variations. Accordingly, a space width in the hybrid resist can be selectively printed by varying the mask-wafer gap distance, allowing more versatile structures to be formed and adjustments to be made for process changes such as resist composition and ion implant levels. |
051679108 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention was made based on the study of the techniques disclosed in JP-A-57-53688 and JP-A-60-207095. In the eddy current sensor disclosed in JP-A-57-53688, a detection accuracy of the sensor is lowered when a groove narrower than a diameter of the eddy current sensor or a hole having a narrower diameter than the diameter of the eddy current sensor is to be detected. Characters of 8 mm square size are engraved in a handle of a fuel assembly. A width of those curved characters is as narrow as approximately 1.6 mm. Accordingly, the detection accuracy of the nuclear fuel identification number by the eddy current sensor is low. Since the diameter of the eddy current sensor is usually approximately 5-10 mm, the reduction of the detection accuracy is inevitable. In the ultrasonic wave sensor disclosed in JP-A-60-207095, a signal processing time for detecting the nuclear fuel identification number is long and it is difficult to apply the sensor to a number of fuel assemblies in a fuel storage pool, which require a short response time. This is also true in a case where an eddy current sensor having a long signal processing time is used. JP-A-57-110994 is similar to JP-A-60-207095. The present invention is intended to solve those problems. One embodiment of the nuclear fuel identification code reader of the present invention is now explained with reference to FIGS. 1, 2 and 3. The nuclear fuel identification code reader of the present embodiment comprises a sensor handling unit 1, an ITV camera 5, an ultrasonic wave probe 9, an ultrasonic wave probe scanner 10, a nuclear fuel handling control unit 22, a nuclear fuel identification code monitor 28 and a nuclear fuel identification code processing unit 43. The sensor handling unit 1 utilizes a portion of a nuclear fuel handling unit 51. The nuclear fuel handling unit 51 is used to move a used fuel assembly 66 in a fuel storage pool 63. The used fuel assembly 66 is taken out of a core of a nuclear reactor and loaded and stored in a fuel rack 65 arranged in the fuel storage pool 63. The fuel storage pool 63 is filled with water 64. The fuel rack 65 is arranged under a water level of the water 64. The nuclear fuel handling unit 51 is constructed to cross the fuel storage pool 63. The nuclear fuel handling unit 51 includes a movable truck 52, a laterally movable truck 53, a grapple 54, a clamp 55 and a hoist 56. The movable truck 52 is driven by a drive motor 58 on a pair of rails 57 arranged on both sides of the fuel storage pool 63. The laterally movable truck 53 has the grapple 54, the clamp 55 and the hoist 56 and is driven by a drive motor 60 on a pair of rails 59 arranged on the movable truck 52. The grapple 54 is raised and fallen by the hoist 56 and a drive motor 62. The grapple 54 is rotated by a drive motor 61 to allow adjustment of an angle of the clamp 55 in a horizontal plane. The drive motor 61 is mounted on the laterally movable truck 53. The grapple 54 is constructed by several linked expandable pipes. The lamp 55 is mounted at the bottom end of the grapple 54. A direction of movement of the movable truck 52 is represented by X, a direction of movement of the laterally movable truck 53 is Y, the elevation of the grapple 54 is represented by Z.sub.1 and the rotation is represented by .theta..sub.1. Position signals representing the displacements X, Y, Z.sub.1 and .theta..sub.1 are detected by synchronous signal generators (not shown) mounted on the respective drive shafts of the fuel handling unit 51. The sensor handling unit 1 has a hoist 2, a drive motor 3 and a grapple 4. The hoist 2, the drive motor 3 and the grapple 4 are mounted on the movable truck 53. A drive motor (not shown) which corresponds to the drive motor 61 and serves to rotate the grapple 4 is mounted on the laterally movable truck 53. The grapple 4 is also constructed by several linked expandable pipes. The grapple 4 is moved up and down by the hoist 2 and the drive motor 3. The elevation of the grapple 4 is represented by Z.sub.2 and the rotation is represented by .theta..sub.2. Position signals representing the displacements Z.sub.2 and .theta..sub.2 are detected by synchronous signal generators (not shown) mounted on the respective drive shafts of the sensor handling unit 1. A mount table 6 is arranged at the bottom end of the grapple 4. The ITV camera 5 is mounted on the mount table 6. Four downwardly extending frames 7 are fixed to the mount table 6 by screws. An illumination device 8 is arranged at the bottom ends of the frames 7. The ultrasonic wave probe scanner 10 is mounted on the mounted table 6 by an arm 11. A detailed structure of the ultrasonic wave probe scanner 10 is explained below. A drive motor 13 is mounted at a top of a bottom-opened box 12. An upwardly extending rotary screw 14 linked to a rotation shaft of the drive motor 13 meshes with a nut (not shown) mounted on a support member 16 which is mounted on the arm 11. A pair of guide members 15A and 15B which hold the rotary screw 14 therebetween are mounted on the drive motor 13. The guide members 15A and 15B extend through the support member 16 so that they are vertically movable. Another drive motor 17 is mounted on a side of the box 12 which faces the ITV camera 5. A rotary screw 18 horizontally arranged in the box 12 has one end thereof linked to a rotation shaft of the drive motor 17 and the other end thereof supported by a bearing (not shown) mounted on the side of the box 12. A probe holding table 19 engages with the rotary screw 18. Two ultrasonic wave probes 9 are mounted on the probe holding table 19. They are arranged to traverse the rotary screw 18. A pair of fixing guides 21 is arranged in the box 12. An encoder 20 measures the displacement of the ultrasonic wave probe 9 along the axis of the rotating screw 18. The ITV camera 5 and the ultrasonic wave probe 9 constitute a nuclear fuel identification code detection means. The nuclear fuel handling control unit 22 comprises input/output means 23A and 23B, a nuclear fuel handling unit control means 24, a nuclear fuel monitor unit control means 25, a memory 26 and a console panel 27. The input/output means 23A supplies control signals to the drive motors 58, 60, 61 and 62 and receives the position signals representing the displacements X, Y, Z.sub.1 and .theta..sub.1 from the corresponding synchronous signal generators (not shown). The input/output means 23B supplies control signals to the drive motors 3, 58 and 60 and the drive motor (not shown) which drives the grapple 4, and receives the position signals representing the displacements X, Y, Z.sub.2 and .theta..sub.2 from the corresponding synchronous signal generators (not shown). The input/ output means 23A inputs and outputs the signals related to the nuclear fuel handling unit control means 24, and the input/output means 23B inputs and outputs the signals related to the nuclear fuel monitor unit control means 25. The nuclear fuel handling unit control means 24 and the nuclear fuel monitor control means 25 are included in a computer 48. The nuclear fuel identification code monitor 28 has a video signal digitizer 29 and a signal processing microprocessor 30. The microprocessor 30 has a memory 33, an image processing means 31 and a nuclear fuel number identification/discrimination means 32. The video signal digitizer 29 is a kind of A/D converter which converts a video signal (analog signal) sent from the ITV camera 5 to a digital signal. The video signal digitizer 29. The image signal processing means 31 and the fuel number identification/discrimination means 32 are coupled to the input/output means 23B. The nuclear fuel identification code monitor 28 thus constructed identifies the nuclear fuel identification code based on the video signal derived from the ITV camera 5. The nuclear fuel identification code monitor 34 comprises a pulse generation means 35, a signal receive means 36, a probe scanner control means 38 which is constructed by a microprocessor, and a signal processing microprocessor 39. The microprocessor 39 has an ultrasonic wave signal processing means 40, a nuclear fuel number identification means 41, and a memory 42. The pulse generation means 35 is connected to the ultrasonic wave probe 9 and the probe scanner control means 38. The signal receive means 36 is connected to the ultrasonic wave probe 9 and the ultrasonic wave signal processing means 40. The ultrasonic wave signal processing means 40 is coupled to the encoder 20. The probe scanner control means 38 is connected to the drive motors 13 and 17, the encoder 20 and a limit switch 37, and further to the input/output means 23B. The nuclear fuel identification code monitor 34 thus constructed identifies the nuclear fuel identification code based on the reflected wave of the ultrasonic wave derived from the ultrasonic wave probe 9. The nuclear fuel identification code processing unit 43 comprises a fuel number processing means 44 and a memory 45. A numeral 46 denotes a display and a numeral 47 denotes a printer. The display 46 may be mounted on the console panel 27. A structure of the fuel assembly 66 loaded in the fuel rack 65 is explained with reference to FIG. 4. The fuel assembly 66 loaded in the fuel rack 65 is an assembly of used fuel which is taken out of a center of a boiled-water type nuclear reactor. The fuel assembly 66 comprises an upper tie plate 67, a lower tie plate 69, a plurality of fuel rods 70 and a plurality of fuel spacers 71. The top and bottom ends of the fuel rods 70 are held by the upper tie plate 67 and the lower tie plate 69. The fuel spacers 71 are arranged axially of the fuel assembly 66 to keep a predetermined spacing between the fuel rods 70. A channel box 72 mounted on the upper tie plate 67 surrounds a bundle of fuel rods 70 supported by the fuel spacers 71. The upper tie plate 67 has a handle 72 arranged at the top thereof. A nuclear fuel identification number 74 is marked on the top 73 of the handle 72. As shown in FIGS. 5 and 6, the nuclear fuel identification number 74 includes a nuclear fuel identification number 74A coded by recesses 75 having circular cross-sections, and a nuclear fuel identification number 74B which is a combination of alphanumeric characters. Those two types of nuclear fuel identification numbers are marked in parallel on the top 73 of the handle 72. The nuclear fuel identification number 74B can be recognized by a human when he/she looks it but the nuclear fuel identification number 74A cannot be recognized by the human by just looking it. Both of the nuclear fuel identification numbers 74A and 74B are marked by engraving on the top 73 of the handle 72. The nuclear fuel identification number 74A is a combination of the recesses 75 which corresponds to the nuclear fuel identification number 74B. The cross-section of the recess 75 need not necessary be circular but it may be trianglar, square or rectangular, or even oval. In the nuclear fuel identification number 74A, each area sectioned by broken lines 76 corresponds to one character. In FIG. 5, the nuclear fuel identification number 74A represents "2FABC". Each coded symbol of the nuclear fuel identification number 74A is represented by the combination of up to six recesses 75 (two lines of three recesses). Each symbol of the nuclear fuel identification number 74A is a digital signal represented by the presence or absence of the recess 75. FIG. 7 shows a correspondence between the digitized symbols representing the nuclear fuel identification number 74A and the alphanumeric characters (0-9, A-Z). In FIG. 7, the black dot represents the presence of the recess 75 and a white dot represents the absence of the recess 75. It is possible to digitize 36 alphanumeric characters by arranging six recesses 75 (two lines of three recesses) as shown in FIG. 7. A plurality of digital symbols shown in FIG. 7 may be arranged on the top 73 in combination with the characters of the nuclear fuel identification number 74B. Assuming that a diameter of the recess 75 is approximately 1 mm and a spacing W.sub.1 between the recesses 75 in one digital symbol is at least approximately 1 mm, the presence or absence of the recess 75 can be detected by the ultrasonic wave. Since a spacing W.sub.2 between lines of recesses 75 is approximately 3 mm, the nuclear fuel identification number 74A and the nuclear fuel identification number 74B can be marked in parallel on the top 73 having a width of approximately 12 mm. A recess 76 which is used as a reference to read the nuclear fuel identification number 74A is formed on the top 73 of the handle 72. The recess 76 is orthogonal to the side of the handle 72 and it is positioned on the left of the first digital symbol of the nuclear fuel identification number 74A. Without the recess 76, the nuclear fuel identification code monitor 34 cannot specify the nuclear fuel identification number 74A detected by the ultrasonic wave. In the example shown in FIG. 5, the nuclear fuel identification number 74A may be read as either "2FABC" or "CBAF2". If it is determined that the nuclear fuel identification number 74A is to be read from the end adjacent to the recess 76, the number 74 is read as "2FABC". A width W of the recess 76 is either wider or narrower than a width W.sub.4 (diameter) of the recess 75 so that the nuclear fuel identification code monitor 34 can easily discriminate the recess 75 of the digital symbol and the recess 76 of the read reference. The recess 76 need not be linear but it may be circular, triangular or square in cross-section so long as it is positioned on the left of the lines of recesses 75. The fuel assembly 66 having the nuclear fuel identification numbers 74A and 74B marked in parallel on the top 73 of the handle 72 is loaded into the center of the boiled-water type nuclear reactor after the used fuel assembly 66 has been removed from the center of the reactor. The operation of the nuclear fuel identification code reader of the present invention is non explained. An operator specifies, through the console panel 27, the operation of the nuclear fuel handling or the operation of the detector for the nuclear fuel identification code. The specified operation signal is supplied to the computer 48. If the former operation is specified, the nuclear fuel handling unit control means 24 is activated, and if the latter operation is specified, the nuclear fuel monitor control means 25 is activated. It is now assumed that the specified operation is the operation of nuclear fuel handling. Before the function of the nuclear fuel handling unit control means 24 is explained, the operation of the nuclear fuel handling unit 51 in the nuclear fuel handling operation is briefly explained. A plurality of used free assemblies 66 are carried to a predetermined position in the fuel storage pool 63 from the top thereof while they are loaded in a container. Then, the movable truck 52 and the laterally movable truck 53 are driven to move the clamp 55 above the container. As the grapple 54 descends, the clamp 55 is lowered to the position of the handle 72 of the fuel assembly 66 in the container. After the clamp 55 has held the handle 72, the grapple 54 is raised. When the bottom end of the fuel assembly 66 reaches a level which is a predetermined distance above the top end of the fuel rack 65, the elevation of the grapple 54 is stopped. The movable truck 52 and the laterally movable truck 53 are again driven to move the fuel assembly 16 to a level which is a predetermined distance (specified by the operator through the console panel 27) above the fuel rack 65. When it reaches that level, the grapple 54 is lowered to load the fuel assembly 66 to a predetermined position in the fuel rack 65. The above movement is referred as a movement 1. When the fuel assembly is taken out of the fuel storage pool 63 for fuel processing, the opposite movement (movement 2) is carried out. Namely, the fuel assembly 66 taken out of the fuel rack 65 is loaded into the container. The clamp 55 may be moved, while it does not clamp the fuel assembly 66, from the position of the fuel rack 65 to other position (movement 3), from the position of the fuel rack 65 to the position of the container (movement 4), or from the position of the container to the position of the fuel rack 65 (movement 5). The memory 26 stores data relating to the loading status of the fuel assembly 66 at the respective positions (X-Y ID coordinate) of the fuel rack 65. The memory 26 stores "0" for the position at which the fuel assembly 66 is not loaded, and "1" for the position at which the fuel assembly 66 is loaded. The ID coordinate is not represented by absolute distances on X and Y axises but it is represented by the code applied to the fuel assembly load position. The position signals representing the displacements X, Y, Z.sub.1 and .theta..sub.1 measured by the synchronous signal generators are converted to the digital signals by the input/output means 23A and they are supplied to the nuclear fuel handling unit control means 24 of the computer 48. A limit switch (not shown) mounted on the nuclear fuel handling unit 51 detects when the grapple 55 reaches a grapple upper limit level A and a mount level B of the fuel assembly (in the fuel rack 65 and the container). The detection signal is supplied to the computer 48. The nuclear fuel handling unit control means 24 uses those signals to control and monitor the position of the nuclear fuel handling unit 51. When the used fuel assembly 66 is to be moved in the fuel storage pool 63, the operator specifies N target positions (X-Y ID coordinate) necessary for the fuel handling unit 51 to move the fuel assembly 66, through the console panel 27. The P.sub.l or P which is shown in JP-B-58-21238, column 6, lines 8-10 is also specified through the console panel 27. The nuclear fuel handling unit control means 24 discriminates in the manner described and shown in JP-B-58-21238, column 7, line 26 to column 8, line 11 and FIGS. 3 and 4. That is, the load status of the fuel assembly 66 at the target position, the correctness of the data (P.sub.l or P) specified by the operator and the open/close status of the clamp 55 are checked, and if the check result is normal, the control signal for the corresponding movement (one of the movements 1-5) is supplied to the nuclear fuel handling unit 51 to control the corresponding movement. If the check result is not normal, the nuclear fuel handling unit control means 24 inhibits the start of the nuclear fuel handling unit 51. When the fuel assembly 66 is moved by the nuclear fuel handling unit 51, the data on the load status of the fuel assembly 66 at the respective positions of the fuel rack 65, which is stored in the memory 26, is updated as the movement proceeds. When the operator specifies the operation of the detector for the nuclear fuel identification code, the nuclear fuel monitor unit control means 25 moves the sensor handling unit 1 in accordance with the process (steps 77A-77M) of FIG. 8. This is explained in detail below. When a monitor operation signal is supplied from the console panel 27 (step 77A), a step 77B is carried out. The position signals representing the displacements X, Y, Z.sub.2 and .theta..sub.1 measured by the respective synchonous signal generators are converted to the digital signals by the input/output means 23B and they are supplied to the nuclear fuel monitor control means 25 of the computer 48. Levels L.sub.1 and L.sub.2 are detected by the limit switch (not shown) mounted on the sensor handling unit 1. Those detection signals are supplied to the nuclear fuel monitor control means 25, which uses those signals to control and monitor the position of the sensor handling unit 1. The level L.sub.1 (FIG. 2) is set at the bottom end of the ITV camera 5 when the nuclear fuel identification number is monitored so that the illumination device 8 does not contact to the top 73 of the handle 72 of the fuel assembly 66 in the fuel rack 65. The level L.sub.2 is set substantially above the level L.sub.1, at a position where the ITV camera 5 is positioned when the nuclear fuel identification number is not detected. The sequence of the fuel assemblies 66 for which the nuclear fuel identification numbers are detected is predetermined and stored in the memory 26. The sequence is shown in FIG. 1 by a chain line 49 starting at a point K.sub.1. It is in the order of the positions K.sub.i (i=1, 2, . . . n.sub.1) shown in the X-Y ID coordinate for the fuel rack 65. In a step 77B, the grapple 4 is lowered, and when the bottom end of the ITV camera 5 reaches the level L.sub.1, the descend of the grapple 4 is stopped. Then, i is set to "1" (step 77C). Whether i=n.sub.1 or not is checked (step 77D). If it is, a step 77E is carried out, and if it is not, a step 77M is carried out. In a step 77F following to the step 77E, the movable truck 52 and the laterally movable truck 53 are driven so that the ITV camera 5 reaches on the fuel assembly 66 which is at the position K.sub.i for which the nuclear fuel identification number is to be detected. When i=1, the ITV camera 5 is set to the start position K.sub.1. When the ITV camera 5 reaches the position K.sub.i, a start of detection signal S.sub.1 for the nuclear fuel identification number 74B by the ITV camera 5 is produced (step 77G). The start of detection signal S.sub.1 is supplied to the nuclear fuel identification code monitor 28, the ITV camera 5 and the illumination device 8. When the nuclear fuel identification code monitor 28 receives the start of detection signal S.sub.1, it starts to receive and process the video signal produced by the ITV camera 5. Upon receipt of the signal S.sub.1, the ITV camera 5 starts to pick up the image and the illumination device 8 is turned on. The image pick-up operation of the ITV camera 5 and the turn-on of the illumination device 8 may be started by the input of the signal S.sub.1 at the position K.sub.1 and continued until the image pick-up at the position K.sub.n is completed, instead of repetitively turning on and off at each position. In a step 77H, a discrimination signal J is received. The discrimination signal J is produced by the nuclear number identification/discrimination means 32 when the processing of the video signal relating to one fuel assembly 66 is completed in the nuclear fuel identification code monitor 28. The fuel member identification/discrimination means 32 produces a "0" discrimination signal J when all characters of the fuel identification number 74B detected by the ITV camera 5 are recognized by the image processing, and produces a "1" discrimination signal J when all characters are not recognized. After the step 77H, whether the discrimination signal J is "1" or "0" is checked (step 77I). If the discrimination signal J is "0", a step 77D is carried out, and if the signal J is "1", a step 77J is carried out. In the step 77J, the laterally movable truck 53 (or the movable truck 52) is driven to move the ultrasonic wave probe 9 on the position K.sub.i. When the ultrasonic wave probe 9 reaches the position K.sub.i, a start of detection signal S.sub.2 for the nuclear fuel identification number 74A by the ultrasonic wave probe 9 is produced (step 77K). The start of detection signal S.sub.2 is supplied to the probe scanner control means 38 of the nuclear fuel identification code monitor 34 to effect the detection of the nuclear fuel identification number 74A by the ultrasonic wave probe 9 and the recognition of the nuclear fuel identification number 74A by the nuclear fuel identification code monitor 34. The probe scanner control means 38 produces an end of ultrasonic wave scan signal E.sub.1 when the scan of the ultrasonic wave probe 9 to detect the fuel identification number 74A is over. When the nuclear fuel monitor control unit 25 receives the end signal E.sub.1 (step 77L), it carries out the decision of the step 77D. If the decision in the step 77D is YES, the grapple 4 is elevated to elevate the ITV camera 5 to the level L.sub.2 (step 77M). In this manner, the nuclear fuel identification numbers of all fuel assemblies 66 in the fuel storage pool 63 are monitored. The nuclear fuel monitor control means 25 drives the ITV camera 5 which is the optical sensor for the nuclear fuel identification number onto the fuel assembly 66 under consideration, and when the fuel identification number 74B detected by the ITV camera 5 is hard to be recognized, the ultrasonic wave probe 9 which is the ultrasonic wave sensor is driven onto the fuel assembly 66 to detect the fuel identification member 74A. The operation of the nuclear fuel identification code monitor 28 when it receives the start of detection signal S.sub.1 produced by the nuclear fuel monitor unit control means 25 is explained. When the start of detection signal S.sub.1 is received, the video signal digitizer 29 starts the A/D conversion of the video signal for the top 73 of the handle 72 picked up by the ITV camera 5. The image signal converted to the digital signal by the video signal digitizer 29 is supplied to the memory 33 in 1/30 second and stored therein. The image processing means 31 carries out the process shown in FIG. 9. Upon receipt of the start of detection signal S.sub.1, the image processing means 31 receives the video signal stored in the memory 33 (step 78A). It extracts the image signal of the nuclear fuel identification number 74B marked on the fuel assembly 66 under consideration, from the input image signal (step 78B). The extracted image signal is processed for noise elimination (step 78C) and contrast enhancement (step 78D). Then, the image signal is binarized to generate character patterns for all characters (n.sub.2, n.sub.2 =5 in the present embodiment) of the detected nuclear fuel identification number 74B (step 78E). Those character patterns are generated as two-dimension character patterns P.sub.jk (l) (j=1-M, k=1-N, l=1-n.sub.2) having M.times.N picture elements. The n.sub.2 generated character patterns are supplied to the fuel number identification/discrimination means 32 in the sequence of the characters of the nuclear fuel identification number 74B (step 78F). When the nuclear fuel number identification/ discrimination means 32 receives the character patterns of the characters of the nuclear fuel identification number 74B, it carries out a process comprising steps 79A-79K shown in FIG. 10. The nuclear fuel number identification/discrimination means 32 receives the n.sub.2 character patterns P.sub.jk (l) (step 79A) and carries out patterns Q.sub.jk (m) from the memory 33. In the present embodiment, 36 standard character patterns Q.sub.jk (m) (m=1-n.sub.3) including 0-9 and A-Z shown in FIG. 11 are stored in the memory 33, and n.sub.3 =36. Those standard character patterns correspond to the engraved characters of the nuclear fuel identification number 74B. In a step 79E, a similarity I(m) between the character patterns P.sub.jk (l) and the n.sub.3 standard character patterns Q.sub.jk (m) are calculated in accordance with a formula (1). ##EQU1## The similarity I(m) calculated in accordance with the formula (1) is 1.0 when the character patterns P.sub.jk (l) and the standard character patterns Q.sub.jk (m) fully match. It does not exceed 1.0. In a step 79F, whether a maximum one (max I(m)) of the I(m) calculated for the character patterns P.sub.jk (l) is larger than a predetermined threshold S or not. When max I(m) is close to 1.0, it indicates that the ITV camera 5 has detected the characters of the nuclear fuel identification number 74 to a sufficient extent to permit the recognition. When max I(m) is around 0.6, it means that the characters of the nuclear fuel identification number 74B cannot be sufficiently detected because of the deposition of soft clad. The threshold S is to be determined by taking the above into account. When max I(m) is smaller than the threshold S (the decision in the step 79F is NO), the "1" discrimination signal J is supplied to the input/output means 23B of the nuclear fuel handling unit control means 22 in order to detect the nuclear fuel identification number 74A by the ultrasonic probe 9 (step 79K). In the step 79F, the necessity of the detection of the nuclear fuel identification number 74A by the ultrasonic wave sensor, that is, the necessity of the movement of the ultrasonic wave probe 9 onto the fuel assembly 66 under consideration is checked. If the decision in the step 79F is YES, the characters of the character patterns P.sub.jk (l) are recognized as the characters corresponding to the standard character patterns Q.sub.jk (m) having the similarity max I(m) (step 79G). If the decision in the step 79H is NO, the steps 79C et seq are repeated. If the decision is YES, a step 79I (output of the "0" discrimination signal J) is carried out. The "0" discrimination signal J is also supplied to the input/output means 23B. Finally, the n characters (2FABC) recognized in the step 79J are supplied to the nuclear fuel number processing means 44 of the nuclear fuel identification code processing unit 43. The character recognition technique carried out by the nuclear fuel number identification/discrimination means 32 is a two-dimension template matching method. The operation and process of the nuclear fuel identification code monitor 34 when the nuclear fuel number identification/discrimination means 32 produces the "1" discrimination signal J in the step 79K and the nuclear fuel monitor control means 25 produces the start of detection signal S.sub.2 are now explained. The process of the probe scan control means 38 is shown in FIG. 12. When it receives the start of detection signal S.sub.2 (step 80A), it sends a drive signal to the drive motor 13 (step 80B). As the drive motor 13 rotates, the rotary screw 14 is rotated and the box 12 which accommodates the ultrasonic wave probe 9 is moved down. When the start of detection signal S.sub.2 is generated, the ultrasonic wave probe scanner 10 has already been located above the fuel assembly 66 under consideration. As a result, the handle 72 of the fuel assembly 66 is inserted between the pair of fixing guides 21 in the descending box 12. When the limit switch 37 contacts to the top 73 of the handle 72, it produces an activation signal. When the probe scanner control means 38 receives the activation signal, it stops the rotation of the drive motor 13. In a step 80D, a start of ultrasonic wave scan signal S.sub.3 is generated. The start signal S.sub.3 is supplied to the pulse generation means 35 and the ultrasonic signal processing means 40. When the pulse generation means 35 receives the start signal S.sub.3, it produces an electrical pulse to cause the ultrasonic wave probe 9 to generate an ultrasonic wave. The ultrasonic wave generated by the ultrasonic wave probe 9 is irradiated to the top 73 of the handle 72. After the step 80D, a drive signal is supplied to the drive motor 17 (step 80E). As the drive motor 17 rotates, the rotary screw 18 is rotated and the probe support table 19 which accommodates the ultrasonic probe 9 is moved from the right to the left in FIG. 2. Since the fixing guides 21 contact to the top 73 of the handle 72, the ultrasonic wave probe 9 is moved laterally while the distance to the top 73 is kept constant. One of the pair of ultrasonic wave probe 9 mounted on the probe support table 19 moves on an extended line of an arrow R.sub.1 (FIG. 5) and the other moves on an extended line of an arrow R.sub.2 which is parallel to the arrow R.sub.1. In the present embodiment, the pair of ultrasonic wave probes 9 can substantially simultaneously detect the two lines of recesses 75 of the nuclear fuel identification number 74A. When the ultrasonic wave probe 9 reaches the end point of scan, it is detected by the encoder 20. The detection signal (position signal of the ultrasonic wave probe 9) of the encoder 20 is supplied to the probe scanner control means 38. When the ultrasonic wave probe 9 reaches the end point of scan, the probe scanner control means 38 stops the drive motor 17 and produces an end of ultrasonic wave scan signal E.sub.1 (step 80E). After the step 80F, it drives the drive motor 13 to elevate the ultrasonic wave probe 9 to a predetermined position (step 80G). Then, the detection of the nuclear fuel identification number 74A by the ultrasonic wave probe 9 is terminated. As described above, the ultrasonic wave probe 9 irradiates the ultrasonic wave to the top 73 of the handle 72 and receives the reflected ultrasonic wave from the top 73. A relationship between a horizontal position of the ultrasonic wave probe 9 driven by the drive motor 17 and the reflected ultrasonic wave is shown in FIG. 13. The reflected wave in FIG. 13 is detected by the ultrasonic probe 9 which is moved on the extended line of the arrow R.sub.1. The ultrasonic wave generated by the ultrasonic wave probe 9 is mostly reflected by the area of the top 73 which has no recess 75, and the reflected ultrasonic wave is received by the ultrasonic wave probe. However, since the bottom of the recess 75 is arcuate as shown in FIG. 6, the ultrasonic wave is scattered in the area of recess 75 and little reflected wave reaches the ultrasonic wave probe 9. Accordingly, the amplitude of the reflected wave is zero in the area of recess 75. Clad may deposit on the top 73 of the handle 72 of the fuel assembly 66 which the fuel assembly 66 is loaded in the center of the nuclear reactor and a portion of the recesses 75 may be covered by the clad. Even if the recesses 75 on the top 73 of the handle 72 of the fuel assembly 66 loaded in the fuel rack 65 is covered by the clad, it is possible to detect the recesses 75 by the ultrasonic wave. This is due to the fact that there is no substantial difference between acoustic impedances of water and water-containing clad (primary constituent is ferric oxide). The reflected wave signal detected by the ultrasonic wave probe 9 is supplied to the signal receive means 36. In the reflected wave signal of FIG. 13, a zero reflected wave output area having a width W.sub.3 corresponds to the recess 76 which is the read reference recess. Other zero reflected wave output areas correspond to the recesses 75. As a method for detecting the digitized recesses 65 of the nuclear fuel identification number 74A, one of the following methods may be adopted: 1) two-dimensionally scanning one ultrasonic wave probe, 2) linearly scanning a plurality of parallelly arranged ultrasonic wave probes, and 3) two-dimensionally scanning the ultrasonic wave beam by electronically switching ultrasonic wave probes by using an array sensor having a plurality of ultrasonic wave probes arranged two-dimensionally to cover the entire area of the nuclear fuel identification number 74A. The present embodiment adopts the method 2). The signal receive means 36 converts the input reflected wave signal to "1" and "0" pulse signals. The zero reflected wave output is converted to "1" and non-zero reflected wave output is converted to "0". The output signal (pulse signal) of the signal receive means 36 and the position signal of the ultrasonic wave probe 9 detected by the encoder 20 are supplied to the ultrasonic wave signal processing means 40, which carries out a process comprising steps 81A-81E shown in FIG. 14. The pulse signal corresponding to the read reference recess 76 detected, and the position of the ultrasonic wave probe 9 where the pulse signal was generated are determined based on the input signals supplied in the step 81A (step 81B). The pulse signal corresponding to the recess 76 can be readily detected because it is narrower (in the area of "1") than the pulse signals corresponding to the recesses 75. The presence or absence of the recesses 75 is detected and the positions of the recesses 75 are determined relative to the recess 76 (step 81C). The pulse signal corresponding to the recess 75, that is, the pulse signal having the pulse width W.sub.4 is detected and the position of the ultrasonic wave probe 9 corresponding to the pulse signal of the pulse width W.sub.4 is determined. Based on the data of the position of the recess 75 determined in the step 81C, the presence or absence of the recess 75 at six predetermined positions is determined in five areas sectioned by the broken lines 76 of the nuclear fuel identification number 74A of FIG. 5, and signals "0" or "1" are assigned to the six predetermined positions, with each predetermined position being a unit (step 81D). Thus, the digital pattern signals for the units, which are "1" if the recesses 75 are present and "0" if the recesses 75 are not present, are produced. A unit number of digital pattern signals corresponding to the number of characters (n.sub.2) of the nuclear fuel identification number 74B are supplied to the fuel number recognition means 41 in sequence starting from the recess 76 (step 81E). The fuel number recognition means 41 carries out a process comprising steps 82A-82C shown in FIG. 15. The memory 42 stores the correspondence between the standard digital patterns which represent the presence or absence of the recesses 75 shown in FIG. 7 and the characters (alphanumeric). The fuel number recognition means 41 reads the standard digital patterns corresponding to the digital patterns for the respective units received in the step 82A, from the memory 42, and recognizes the characters corresponding to the standard digital patterns as the characters for the digital pattern signals (step 82B). The fuel number recognition means 41 supplies the n.sub.2 characters (2FABC) recognized for the fuel identification number 74A to the fuel number processing means 44 (step 82C). Thus, the detection of the nuclear fuel identification number marked on the handle 72 of the fuel assembly 66 by the nuclear fuel identification code monitor 28 or 34, and the recognition of the detected nuclear fuel identification number as characters are terminated. The fuel number processing means 44 receives the characters of the nuclear fuel identification number recognized by the nuclear fuel identification code monitor 28 or 34, and the X-Y ID coordinates of the positions K.sub.i based on the values X and Y inputted to the nuclear fuel monitor control means 25. The fuel number processing means 44 stores the recognized characters of the nuclear fuel identification number and the X-Y ID coordinates of the positions K.sub.i in an associated manner, and displays them on the display 46 and prints them out by the printer 47. Since the recognized characters of the nuclear fuel identification number and the X-Y ID coordinates are associated, the nuclear fuel identification number of the fuel assembly 66 loaded at the position K.sub.i of the fuel rack 65 in the fuel storage pool 63 can be readily determined. In accordance with the nuclear fuel identification code reader of the present embodiment, the following advantages are offered. Since the nuclear fuel identification number 74B marked by the letters is recognized by the optical sensor and the nuclear fuel identification code monitor 28, the fuel assembly 66 in the fuel storage pool can be identified in a short time. Even if it is difficult to recognize the letters of the nuclear fuel identification number 74B based on the video signal from the optical sensor (due to the deposition of the clad to the handle 72 of the fuel assembly 66 under consideration), it is possible to readily recognize the letters of the nuclear fuel identification number by the ultrasonic wave sensor and the nuclear fuel identification monitor 34. By the combined use of the detection of the primary nuclear fuel identification number by the optical sensor and the detection of the supplementary nuclear fuel identification number by the ultrasonic wave sensor, the nuclear fuel identification numbers marked on all fuel assemblies in the fuel storage pool 63 can be detected in a very short time with an accuracy of essentially 100% (99.99%). By preferentially using the detection by the optical sensor to the fuel assembly 66 and supplementarily using the detection by the ultrasonic wave sensor in case the letters of the nuclear fuel identification number 74B cannot be recognized based on the information derived from the optical sensor, the above advantages, particularly the reduction of the detection time, are remarkable. In the present embodiment, for those of the fuel assemblies 66 stored in the fuel storage pool 63 whose nuclear fuel identification number 74B cannot be recognized by the optical sensor, the detection of the nuclear fuel identification number 74A by the ultrasonic wave sensor is effected. The automatic reading of the nuclear fuel identification number may also be effected. In the present embodiment, since the ultrasonic wave sensor detects the digitized recesses 75 formed on the top 73 of the handle 72 of the fuel assembly 66, the processing time for recognizing the letters can be significantly reduced compared to that required in detecting the letters themselves by the ultrasonic sensor. In the present embodiment, the structure of the ultrasonic wave probe scanner and the structure of the associated nuclear fuel identification code monitor (especially a processing program) can be simplified compared to a case where the letters themselves are detected by the ultrasonic wave sensor. The provision of the read reference recess 76 on the top 73 of the handle 72 facilitates the recognition of the letters of the nuclear fuel identification number 74A based on the reflected ultrasonic wave. Since both the digitized (coded) nuclear fuel identification number 74A and the nuclear fuel identification number 74B in letters are marked on the top 73 of the handle of the fuel assembly 66, the detection by the optical sensor and the detection by the ultrasonic wave sensor are facilitated. The provision of the nuclear fuel identification number 74B also permits visual recognition by a human. Since the sensor handling unit 1 is provided in the nuclear fuel handling unit 51, a portion of the nuclear fuel handling unit 51 can be utilized as the nuclear fuel identification code reader and the entire construction can be very compact. In other words, the fuel assembly may be moved by the nuclear fuel identification code reader. In FIG. 1, separate movable truck and laterally movable truck such as grapple 54 for handling the fuel may be provided, although the construction may be complex. The nuclear fuel handling unit control means 24 may also be assembled in a separate computer. The mounting of the ITV camera 5 and the ultrasonic wave probe scanner 10 on one grapple 4 significantly contributes to the simplification of the structure. Since the drive mechanism (the drive motor 13 and the rotary screw 14) for moving up and down the ultrasonic wave probe 9 is provided separately from the grapple 4, the positioning of the ultrasonic probe 9 above the fuel assembly is facilitated. Since the nuclear fuel identification number 74A including the recesses 65 is marked on the top 73 of the fuel assembly 66, the processing time required for the character recognition based on the reflected ultrasonic wave is essentially same as the processing time required for the character recognition based on the detection of the nuclear fuel identification number 74B by the ITV camera 5. However, the detection of the nuclear fuel identification number 74B by the ITV camera 5 can be continuously effected while the movable truck 52 and the laterally movable truck 53 are driven, but the detection of the nuclear fuel identification number 74A by the ultrasonic sensor should repeat the start and stop of the movement of the movable truck 52 and the laterally movable truck 5 and the start and stop of the scan of the ultrasonic wave probe 9 for each fuel assembly 66. Accordingly, the time required to recognize the nuclear fuel identification numbers of all fuel assemblies is shorter when both the character recognition for the fuel identification number 74B based on the video signal from the optical sensor and the character recognition based on the reflected wave by the ultrasonic wave sensor are used than when the characters of the nuclear fuel identification number 74A are recognized by the ultrasonic wave sensor. Another embodiment of the nuclear fuel identification code reader of the present invention is explained with reference to FIGS. 16 and 17. The like elements to those shown in the embodiment of FIG. 1 are designated by the like numeral. Most elements of the present embodiment are identical to those of the embodiment of FIG. 1. In the present embodiment, the nuclear fuel identification code processing unit 43 in the embodiment of FIG. 1 is replaced by a nuclear fuel identification code compare unit 83. Configuration and operation of the nuclear fuel identification code compare unit 83 are explained below. The nuclear fuel identification code compare unit 83 has a memory 45 and fuel number compare means 84. The fuel number compare means 84 receives the letters of the nuclear fuel identification number recognized by the nuclear fuel identification code monitor 28 or 34, and also receives an X-Y ID coordinate of a position K.sub.i based on the values X and Y inputted to the nuclear fuel monitor control means 25 (step 85A). The memory 45 stores the nuclear fuel identification number (in letters) of the fuel assemblies 66 at each position K.sub.i of the fuel rack in the fuel storage pool 63. Those are previously detected data. The memory 45 also stores data representing the load status of the fuel assembly 66 fetched from the memory 26 by the fuel number compare means 84. This data indicates the presence or absence of the fuel assembly 66 at each position K.sub.i of the fuel rack 65. The fuel number compare means 84 compares the X-Y ID coordinate of the position K.sub.i inputted currently and the letters of the nuclear fuel identification number for the position K.sub.i with the corresponding past data read from the memory 45 (step 85B), and determines the matching (step 85C). The comparison result is stored in the memory 45, displayed on the display 46 and printed out by the printer 47. If the result in the result in the step 85C is non-match, a buzzer is sound to alarm to the operator. In this manner, the present embodiment can attain the same advantages as those of the embodiment of FIG. 1, and it is particularly effective in the recognition of the fuel assembly 66 where the fuel assembly 66 is to be stored in the fuel storage pool 6 for an extended period. By the comparison of the letters of the current nuclear fuel identification number and the past data previously detected, the storage of the fuel assembly 66 having the identical nuclear identification number in the fuel storage pool 63 can be readily checked. Other embodiment of the nuclear fuel identification code reader of the present invention is explained below. As shown in FIG. 18, the present embodiment uses a nuclear fuel identification code monitor 28A in place of the nuclear fuel identification code monitor 28 of FIG. 1. The nuclear fuel identification code monitor 28A comprises, in addition to the elements of the nuclear fuel identification code monitor 28, ITV camera control means 86 which receives a start of detection signal S.sub.1 from the nuclear fuel monitor control means 25. The ITV camera control means 86 generates a start of image pickup trigger signal to the ITV camera 5, a turn-on signal to the illumination unit 8, a start of input signal for the video signal to the video signal digitizer 29, and a start of image processing signal to the image processing means 31, in response to the input start of detection signal S.sub.1. When the ITV camera 5, the illumination unit 8, the image signal digitizer 29 and the image processing means 31 receives those signals, they carry out the functions assigned thereto as they do in the embodiment of FIG. 1. The present embodiment also attains the same advantages as those of the embodiment of FIG. 1. A software implemented embodiment of the image processing means 31 and the fuel number identification/ discrimination means 32 of the microprocessor 30 used in the nuclear fuel identification code monitor 28 shown in FIG. 1 is explained with reference to FIG. 19. A microprocessor 30A which corresponds to the microprocessor 30 has input means 87A, output means 87B, a CPU 87C, a RAM 87D and a ROM 87E. The input means 87A is connected to the video signal digitizer 29 and the input output means 23B. The output means 87B is connected to the input/output means 23B and the fuel number processing means 44. An internal bus 87F connects the input means 87A, the output means 87B, the CPU 87C, the RAM 87D and the ROM 87E in the microprocessor 30A. The function of the memory 33 of the microprocessor 30 is effected by the RAM 87D. The output of the video signal digitizer 29 is stored in the RAM 87D. The ROM 87E stores the processing steps shown in FIGS. 9 and 10 with the steps 78F and 79A being removed and the step 79B being executed after the step 78E. In the present embodiment, the memory 33 of FIGS. 9 and 10 is replaced by the RAM 87D. The CPU 87C recognizes the letters of the nuclear fuel identification number 74B based on the video signal from the ITV camera 5 in accordance with the process stored in the ROM 87E. A software implemented embodiment of the ultrasonic signal processing means 40 and the fuel number identification means 41 of the microprocessor 39 used in the nuclear fuel identification code monitor 34 of FIG. 1 is explained with reference to FIG. 20. A microprocessor 39A corresponding to the microprocessor 39 has input means 89A, output means 89B, a CPU 89C, a RAM 87D and a ROM 89E which are interconnected through an internal bus 89F. The input means 89A is connected to the encoder 20 and the probe scanner control means 38. The output means 89B is connected to the fuel number processing means 44. The function of the memory 42 of the microprocessor 39 is effected by the RAM 87D. The processing steps shown in FIGS. 14 and 15 with the steps 81E and 82A being removed and the step 82B being executed after the step 81D are stored in ROM 87E. The CPU 87C recognizes the letters of the nuclear fuel identification number 74A based on the reflected ultrasonic wave in accordance with the process stored in the ROM 89E. In the embodiment of FIG. 1, the same advantages are attained when the microprocessor 30 is replaced by the microprocessor 30A and the microprocessor 39 is replaced by the microprocessor 39A. The recesses 75 of the nuclear fuel identification number 74A marked on the top 73 of the handle 72 of the fuel assembly 66 shown in FIG. 5 may be of an upside-down conical shape at the bottom as shown in FIG. 21. Preferably, the bottom of the recess 75 does not have a flat area which is parallel to the top 73. With such a shape, the scatter of the ultrasonic wave radiated to the recess 75 is violent and the reflected wave from the recess 75 back to the ultrasonic wave probe 9 is very little. As a result, the detection of the recess by the ultrasonic wave is facilitated. The technical concept of the above embodiments may be utilized in recognizing the nuclear fuel identification number marked on a fuel assembly of a pressured water type nuclear reactor. Other embodiment of the nuclear fuel identification code reader of the present invention is explained with reference to FIG. 22. Unlike the above embodiments, the nuclear fuel identification code reader of the present embodiment can be applied to a fuel assembly having no nuclear fuel identification code 74A marked on the top 73 of the handle 72. It can also recognize the nuclear fuel identification code 74B marked on the top 73 of the handle 72 by any one of the output signals of the optical sensor and the ultrasonic wave sensor. The constructions of the nuclear fuel handling unit 51 and the sensor handling unit 1 of the present embodiment are identical to those of the embodiment of FIG. 1. In the present embodiment, the nuclear fuel handling control unit 22, the nuclear fuel identification code monitors 28 and 34, and the nuclear fuel identification code processing unit 43 in FIG. 1 are replaced by a nuclear fuel handling control unit 100, nuclear fuel identification code monitors 250 and 260 and a data processing unit 240, respectively. The nuclear fuel handling control unit 100 has input/output means 23A and 23B, a nuclear fuel handling unit control means 24 and nuclear fuel detection unit control means 101. The input/output means 23A and 23B input and output signals similar to those for the nuclear fuel handling control unit 22 between the nuclear fuel handling unit 51 and the sensor handling unit 1. The nuclear fuel handling unit control means 24 and the nuclear fuel detection unit control means 101 are included in a computer 48A. The nuclear fuel identification code monitor 250 comprises image processing means 140, a video signal digitizer 150, a video frame memory 160, fuel number identification/discrimination means 170 and illumination control means 180. The functions of the image processing means 140, the image frame memory 160 and the fuel number identification/discrimination means 170 are effected by a microprocessor 30A. The illumination control means 180 may also be constructed by the microprocessor. The video signal digitizer 150 has the same function as the video signal digitizer 29. The fuel number identification/discrimination means 170 is coupled to the input/output means. The nuclear fuel identification code monitor 250 reads the nuclear fuel identification code by the video signal produced by the ITV camera 5. The nuclear fuel identification code monitor 260 comprises a signal processing microprocessor 39A, ultrasonic wave scanner control means 190 and ultrasonic wave transmit/receive means 200. The microprocessor 39A has the functions of the ultrasonic wave signal processing means 210 and the fuel number identification means 200A. The ultrasonic wave scanner control means 190 may also be constructed by the microprocessor. The ultrasonic wave transmit/receive means 200 comprises pulse generation means 35 and signal receive means 36. The pulse generation means 35 is connected to the ultrasonic wave probe 9 and the ultrasonic wave scanner control means 190. The signal receive means 36 is connected to the ultrasonic wave probe 9 and the ultrasonic wave signal processing means 210. The ultrasonic wave signal processing means 210 is coupled to the encoder 20 and the fuel number identification means 200. The ultrasonic wave scanner control means 190 is connected to the drive motors 13 and 17, the encoder 20 and the limit switch 37, and also to the input/output means 23B and the fuel number identification means 200A. The fuel number identification means 200 is coupled to the input/ output means 23B and the fuel number identification means 170. The nuclear fuel identification code monitor 260 thus constructed recognizes the nuclear fuel identification code based on the reflected ultrasonic wave from the ultrasonic wave probe 9. The data processing unit 100 comprises fuel number processing means 44, a memory 45 and overall control means 241. The fuel number control means 44 is connected to the display 46, the printer 47, the fuel number identification/discrimination means 170 and the fuel number identification means 200A. The overall control means 241 is connected to the input/output means 23A and 23B, the fuel number identification/discrimination means 170 and the fuel number identification means 200. The memory 45 is connected to the fuel number processing means 44 and the overall control means 241. The console panel 270 is connected to the overall control means 241. The display 46 and the printer 47 may be arranged on the console panel 270. The fuel assemblies 66A (BWR fuel assemblies) whose nuclear fuel identification numbers are to be read by the present embodiment are loaded in the fuel rack 65. The fuel assembly 66A is of the same structure as the fuel assembly 66. However, unlike the fuel assembly 60, the fuel assembly 66A has no nuclear fuel identification number 74A marked on the top 73 of the handle 72. The recess 76 and engraved nuclear fuel identification number 74B are marked on the top 73 of the fuel assembly 66A. The recess 76 has the same function as that in the embodiment of FIG. 1. The operation of the nuclear fuel identification code reader of the present embodiment is now explained. The operator designates the activation of one of the nuclear fuel handling operation and the nuclear fuel identification code detection unit through the console panel 270. The designated activation signal is supplied to the computer 48A. If the former activation signal is designated, the nuclear fuel handling unit control means 24 is activated, and if the latter activation signal is designated, the nuclear fuel detection unit control means 101 is activated. It is assumed that the designated operation is the nuclear fuel handling operation. The nuclear fuel handling activation signal designated by the operator through the console panel 270 is supplied to the nuclear fuel handling unit control means 24 through the overall control means 241 and the input/output means 23A. Then, the nuclear fuel handling unit 51 is controlled by the nuclear fuel handling unit control means 24 as it is done in the embodiment of FIG. 1. Like the memory 26, the memory 45 stores the data relating to the load status of the fuel assembly 66A at each position in the fuel rack 65. The data in the memory 45 is updated when the load status of the fuel assembly 66A in the fuel rack 65 is changed by the movement of the used fuel assembly 66A by the nuclear fuel handling unit 51. The data is updated by the overall control means 241, which receives the related information from the nuclear fuel handling unit control means 24. When the activation of the detection unit is designated by the operator through the console panel 270, the overall control means 241 produces the detection unit activation signal. This signal is supplied to the nuclear fuel detection unit control means 101 through the input/output means 23B. The overall control means 241 reads the positions of the fuel assemblies 66A (positions K.sub.i on the chain line 49 starting at point K.sub.1) whose nuclear fuel identification numbers are to be detected, from the memory 45 and sequentially supplies them to the input/output means 23B at a predetermined time interval. The nuclear fuel detection unit control means 101 receives those signals and controls the movement of the sensor handling unit 1 in accordance with a process shown in FIG. 24. The process shown in FIG. 24 is essentially identical to the process shown in FIG. 8. The process shown in FIG. 24 is different from the process shown in FIG. 8 in that a step 77N is executed after the step 77F, and a step 77P is executed after the step 77J. The step 77N outputs the position V.sub.i of the ITV camera 5 determined based on the position signal representing the displacements X and Y measured by the synchronous signal generators. The step 77P outputs a position W.sub.i of the ultrasonic wave probe 9 determined based on the position signal representing the measured displacements X and Y. The position signal W.sub.i is produced when the fuel number identification means 170 produces a "1" output signal J (which is produced when all letters of the fuel identification number 74B are not recognized). Signals representing the positions V.sub.i and W.sub.i are produced by the input/output means 23B and supplied to the fuel number identification means 170 and the fuel number identification means 200A. Like the nuclear fuel monitor control means 25, the nuclear fuel detection unit control means 101 drives the ITV camera 5 which is the optical sensor for the nuclear fuel identification number onto the fuel assembly 66A under consideration, and drives the ultrasonic wave probe 9 which is the ultrasonic wave sensor onto the fuel assembly 66A for the detection of the fuel identification number 74B when it is difficult to recognize the fuel identification number 74B detected by the ITV camera 5. Through the step 77F, the movable truck 52 and the laterally movable truck 53 are driven and the ITV camera 5 is first moved toward the position K.sub.1. The fuel number identification means 170 receives the positions V.sub.i (X-Y ID coordinate) of the ITV camera 5 which are supplied from time to time by the input/output means 23B through the step 77M. A portion of the process of the fuel number identification/discrimination means 170 is explained with reference to FIG. 27. The fuel number identification/discrimination means 170 inputs, in a step 79L, the predetermined position K.sub.i (initially K.sub.1) supplied by the overall control means 241. In a step 79M, the position V.sub.i is inputted. The position V.sub.i is compared with the position K.sub.i (step 79N). If the decision in the step 79N is YES, it means that the ITV camera 5 is on the fuel assembly 66A which is at the predetermined position K.sub.i. At this point, the fuel number identification/discrimination means 170 supplies the illumination unit turn-on signal to the illumination control means 180 although it is not shown in FIG. 27. The illumination control means 180 turns on the illumination unit 8 in response to this signal. The illumination unit turn-on signal is produced only when the initial position V.sub.1 matches to the initial predetermined position K.sub.1. Then, the illumination unit 8 is kept turned on until the reading of the nuclear fuel identification codes for a predetermined number of fuel assemblies 66A is completed. The fuel number identification/discrimination means 170 produces a start of A/D conversion signal through the step 79P substantially simultaneously with the output of the illumination unit turn-on signal. The video signal digitizer 150 starts the A/D conversion of the video signal for the top 73 of the handle 72 imaged by the ITV camera 5. The image signal converted to the digital signal by the video signal digitizer 29 is supplied to the image frame memory 160 in 1/30 second and stored in the image frame 160. The image processing means 140 executes the process in accordance with the steps shown in FIG. 25. The process shown in FIG. 25 includes steps 78G-78I in addition to the steps shown in FIG. 9. After the steps 78A-78C, the step 78D is carried out. In the step 78D, one of intensity levels 0-255 is assigned to each of a number of pixels corresponding to one character (MXN pixels) in accordance with an image signal level of the corresponding portion. The rank 0 is darkest and the rank 255 is brightest. In the step 78G, a frequency distribution of the rank (FIG. 26) is determined in accordance with the intensity rank of each pixel. The frequency distribution is a distribution of number of pixels having the same intensity rank. A difference between a minimum frequency distribution and a maximum frequency distribution is compared with a predetermined value (step 78H). If the difference is not greater than the predetermined value, it means that the status in the binarization is not good, and a change of intensity signal is produced (step 78I). The illumination control means 180 receives the change of intensity signal to increase the intensity of the illumination unit 8. The time required for processing the above can be reduced to approximately 0.1 second by using a high speed illumination method such as a stroboscope illumination. After the intensity has been changed, the nuclear fuel identification number 74B is picked up by the ITV camera 5. If the decision in the step 78H is YES, the image signal is binarized in the step 78E to prepare the character pattern of the character. The binarization of the image signal is effected by selecting "1" for the signal which is larger than a predetermined reference between the minimum intensity rank and the maximum intensity rank of FIG. 26, and selecting "0" for the signal which is smaller than the predetermined reference. Then, the step 78F is carried out. After the step 78F, the fuel number identification/discrimination means 170 sequentially carries out the steps 79A-79K shown in FIG. 10 as shown in FIG. 27. If the decision in the step 79F is NO, a "1" output signal J is produced in the step 79K as it is in the previous embodiment. The "1" output signal J means that the discrimination of the nuclear fuel identification number 74B by the ITV camera 5 is impossible due to the deposition of the soft clad. The "1" output signal J is also a signal to request redetection of the nuclear fuel identification number 74B by the ultrasonic wave probe 9. When the fuel number identification/discrimination means 170 produces the "1" output signal J, the nuclear fuel detection unit control means 101 carries out the step 77J. The fuel number identification means 200 of the nuclear fuel identification code monitor 260 carries out the steps 76L-79P shown in FIG. 27. The predetermined position K.sub.i is inputted (step 79L). Then, the position W.sub.i of the ultrasonic wave probe 9 which varies from time to time and is produced by the nuclear fuel detection unit control means 101 in the step 77N is inputted. If the position W.sub.i matches to the predetermined position K.sub.i, the start of detection signal S.sub.2 is produced. The ultrasonic wave scan unit control means 190 carries out the steps 80B-80G shown in FIG. 12 when it receives the start of detection signal S.sub.2. The ultrasonic wave probe scan unit 10 is controlled by the signal which is derived through the steps 80B, 80C and 80E-80G. The signal receive means 36 receives the reflected wave signal detected by the ultrasonic wave probe 9. The reflected wave is supplied to the ultrasonic wave signal processing means 210. The ultrasonic wave signal processing means 210 carries out the process shown in FIG. 28. In a step 81A, the reflected ultrasonic wave signal and the position signal of the ultrasonic wave probe 9 detected by the encoder 20 are inputted. The binarization of the reflected wave signal in a step 81F is effected based on a time difference between the reflected wave signals of the focused ultrasonic wave beam radiated from the ultrasonic wave probe 9. The focused ultrasonic wave beam from the ultrasonic wave probe 9 is radiated normally to the top 73 of the handle 72 and the bottom 91 of the nuclear fuel identification number 74B (FIG. 29). The time at which the reflected wave signal for the top 73 is represented by t.sub.1 (FIG. 30A), and the time at which the reflected wave signal for the bottom 91 is received is represented by t.sub.2 (FIG. 30B). A time to represent the time at which the focused ultrasonic wave beam is radiated. A time t.sub.s (=(t.sub.2 -t.sub.1)/2) is set as a threshold level. If the reflected ultrasonic wave signal is detected at a time t where t.sub.s >t, "0" is assigned to the position of the ultrasonic wave probe 9 at which the signal is detected. If the reflected ultrasonic wave signal is detected at a time t where t.sub.s .ltoreq.t, "1" is assigned to the position of the ultrasonic wave probe 9 at which the signal is detected. In this manner the reflected wave signal is binarized. In the present embodiment, the method 2) is used as it is in the previous embodiment. In the present embodiment, three or more ultrasonic wave probes 9 are arranged in parallel. In a step 81G, character patterns of the characters of the nuclear fuel identification number 74B are prepared in accordance with the position signal of the ultrasonic wave probe 9 and the binary signal produced in the step 81F. The prepared character patterns are supplied to the fuel number identification means 200 (step 81H). The fuel number identification means 200 carries out the steps shown in FIG. 31 which are essentially same as the steps shown in FIG. 10. If the decision in the step 79F is NO, the process is terminated. The fuel number processing means 44 stores the characters of the nuclear fuel identification code 74B identified by the nuclear fuel identification monitor 250 or 260 and the corresponding X-Y ID coordinates of the positions K.sub.i in the memory 45. It also displays them on the display 46 and prints them out by the printer 47 as may be required. The present embodiment also attains the same advantages as those of the embodiment of FIG. 1. However, since the nuclear fuel identification number 74B is detected by the ultrasonic wave probe 9 in the present embodiment, a longer recognition time is required than a case where the nuclear fuel identification number 74A is detected. Other embodiment of the nuclear fuel identification code reader of the present invention is explained with reference to FIGS. 32 and 33. In the present embodiment, a Chelencoff light camera 93 and a nuclear fuel monitor 280 are added to the embodiment shown in FIG. 22. The Chelencoff light camera 93 is mounted on the mount table 6 (FIG. 2) of the sensor handling unit 1. The nuclear fuel monitor 280 has a video signal digitizer 110 and a signal processing microprocessor 94 as shown in FIG. 33. The microprocessor 94 has an image frame memory 120, nuclear fuel data processing means 130 and image processing means 140A. The video signal digitizer 110 is coupled to the Chelencoff light camera 93. The nuclear fuel data processing means 130 is coupled to the input/output means 23B, the fuel number processing means and the overall control means 241. The present embodiment has means for determining if the used fuel assembly 66A stored in the fuel storage pool 63 is a real fuel assembly which contains the nuclear fuel. This means comprises the Chelencoff light camera 93 and the nuclear fuel monitor 280. The Chelencoff light camera 93 detects only a light in an ultraviolet range (Chelencoff light) emitted in water by a gamma ray emitted from a nuclear fission seed, amplifies it by a photo-multiplier and directs the amplified electrons to a phosphor plane to visualize them. The image picked up by the Chelencoff light camera 93 has a high intensity at an area corresponding to the water region surrounded by the fuel rods which contain the nuclear fuel. The nuclear fuel detection unit control means 101A of the nuclear fuel handling control unit 100 controls the movement of the sensor handling unit 1 in accordance with a process shown in FIG. 34. In the process shown in FIG. 34, steps 77Q and 77R are added between the steps 77N and 77H of the process shown in FIG. 24. After the movement of the ITV camera 5 in the step 77F, the Chelencoff light camera 93 is moved to the position K.sub.i in the step 77Q. The position U.sub.i of the Chelencoff light camera 93 thus changes from time to time, and the position U.sub.i is outputted (step 77R). The position U.sub.i outputted by the input/output means 23B is supplied to the nuclear fuel data processing means 130. The nuclear fuel data processing means 130 carries out the steps 79L-79P shown in FIG. 27 and supplies the start of A/D conversion signal to the video signal digitizer 110. The video signal digitizer 110 starts the A/D conversion of the video signal supplied from the Chelencoff light camera 93, in response to the start signal. The video signal converted to the digital signal is stored in the image frame memory 120 as the image signal. The image picked up by the Chelencoff light camera 93 does not always have a high S/N ratio. When it is to be determined whether the nuclear fuel is contained in the fuel assembly 66A, no special image processing is required for the video signal picked up by the Chelencoff light camera 93. However, in order to determine whether the nuclear fuel is contained or not in the fuel assembly to prepare a Chelencoff light pattern, the following processing is required. This image processing is done by the image processing means 140A. FIG. 35 shows the image processing means. An image signal is supplied from the image frame memory 120 (step 78A). A plurality of frame images taken time-serially are combined for each pixel (step 78G). A noise component of the image signal is reduced by the combination and the S/N ratio of the image signal is enhanced. The image signal produced in the step 78G is vignetted (step 78H). Through this step, an RF noise component is eliminated from the image signal. The image signal having the RF noise component eliminated is binarized with a proper binarization level (step 78I). The binary data is supplied to the nuclear fuel data processing means 130. The Chelencoff light pattern is created based on the binary data (step 79Q). FIG. 37 shows the Chelencoff light pattern created in the step 78J. In FIG. 37, numeral 70A denotes a fuel rod and numeral 72A denote a handle. The fuel rods which contain the nuclear fuel are distinguished from other elements such as water rods which do not contain the nuclear fuel, and they are patterned. Based on the Chelencoff light pattern thus created, whether the fuel assembly 66A under consideration is the true fuel assembly which contains the nuclear fuel or not is determined (step 79R). The decision and the Chelencoff light pattern are supplied to the fuel number processing means (step 79S). The fuel number processing means 44 carries out the same steps as those of the fuel number processing means 44 of the embodiment shown in FIG. 22, as well as the following steps. If the decision in the step 79R is "true fuel assembly", a reference Chelencoff light pattern corresponding to the nuclear fuel identification number 74B recognized by the nuclear fuel identification code monitor 260 or 280, and the Chelencoff light pattern created in the step 79Q are compared. Through the comparison, it is determined whether the recognized nuclear fuel identification number 74B is correct or not. The result of this determination and the decision in the step 79R are displayed on the display 46. The present embodiment can attain the same advantages as those of the embodiment of FIG. 22. In the present embodiment, it is determined whether the fuel assembly 66A whose nuclear fuel identification number 74B is to be detected is true fuel assembly which contains the nuclear fuel or not. In the present embodiment, the correctness of the nuclear fuel identification number 74B recognized based on the Chelencoff light pattern can be checked and the accuracy of recognition of the nuclear fuel identification number is improved. In the embodiment shown in FIG. 32, since the Chelencoff light camera 93 is driven directly above the handle 72 of the fuel assembly 66A under consideration, the image of the handle 72 is patterned as shown in FIG. 37. As a result, the pattern of the fuel rods located directly below the handle 72 is not created. This problem may be solved by inclining the Chelencoff light camera 93 around the handle 72 by an angle .theta. on each side and picking up the top of the fuel assembly 66A from two directions G.sub.1 and G.sub.2, as shown in FIG. 38. The Chelencoff light camera 93 may be rotated in a direction 95 by a motor (not shown) mounted on the mount table 6. The video signals picked up by the Chelencoff light camera 93 from the directions G.sub.1 and G.sub.2 are supplied to the nuclear fuel monitor 280. Those video signals are converted to digital signals by the video signal digitizer 110 and they are stored in the image frame memory 120. The image processing means 140A processes those image signals for the directions G.sub.1 and G.sub.2 to produce binary data. In a step 79Q of the nuclear fuel data processing means 130, the Chelencoff light patterns created based on the video signals picked up from the directions G.sub.1 and G.sub.2 are combined to create a new Chelencoff light pattern (FIG. 39A). The Chelencoff light pattern created based on the data of the direction G.sub.1 is shown in FIG. 39B, and the Chelencoff light pattern created based on the data of the direction G.sub.2 is shown in FIG. 39C. The Chelencoff light pattern of FIG. 39A is created by image combination of a triangular pattern at right bottom of FIG. 39B and a triangular pattern at left top of FIG. 39C. In the Chelencoff light pattern of FIG. 39A, the handle shown in FIG. 39B and 39C (in broken lines) disappears. Accordingly, the Chelencoff light pattern located directly below the handle is created. The above embodiments are designed to read the nuclear fuel identification number marked on the BWR fuel assembly 66. The PWR fuel assembly 66B has the nuclear fuel identification number marked on the side of the top tie plate. A structure of an optical sensor which can read the nuclear fuel identification numbers marked on both types of fuel assemblies is shown in FIGS. 40A and 40B. FIG. 40A shows a read status for nuclear fuel identification number 74B marked on the fuel assembly 66. FIG. 40B shows a read status for the fuel assembly 66B. When this optical sensor is applied to the fuel assembly 66, a reflection mirror 96 supported by the ITV camera 5 is placed in parallel to the axis of the ITV camera 5. When it is applied to the fuel assembly 66B, the frame 7 and the illumination unit 8 are removed and a reflection mirror tube 97 is mounted on the mount table 6 instead, as shown in FIG. 40B. The reflection mirror tube 97 has a pair of reflection mirrors 99 at the top and the bottom thereof, and has an illumination unit 98 mounted at the lower end. The reflection mirror 96 is rotated to be obliquely to the axis of the ITV camera 5. The lower end of the reflection mirror tube 97 is inserted between the stored fuel assemblies 66B. The image of the nuclear fuel identification number 74B marked on the side at the upper end of the fuel assembly 66B is directed to the ITV camera 5 through the pair of reflection mirrors 99 and the reflection mirror 96. The optical sensor shown in FIGS. 40A and 40B is applicable to the nuclear fuel identification code readers of the above embodiments. |
summary | ||
047740485 | claims | 1. In a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of said vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein said magnetic coil is wound around said toroidal field coil such that said ohmic heating coil encloses said toroidal field coil. 2. The ohmic heating coil system of claim 1 comprising a plurality of said ohmic heating coils. 3. In a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of said vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein said magnetic coil is disposed around the outer perimeter of said toroidal field coil and wherein said toroidal field coil is D-shaped and said magnetic coil is horseshoe-shaped, the opening in said horseshoe-shaped coil being aligned with the first portion of said D-shaped toroidal field coil. 4. An interlocking tokamak reactor module comprising: a toroidal vacuum vessel section; at least one toroidal field coil disposed concentrically around the minor radius of said vacuum vessel section; modular saddle coil means for generating a poloidal confining magnetic field, said saddle coil means disposed circumferentially around the major radius of said toroidal vacuum vessel section and encircling said toroidal field coil about the major radius of said toroidal field coil; and modular ohmic heating coil means. a plurality of interlocking modules, said modules including a toroidal vacuum vessel section; at least one toroidal field coil disposed concentrically around the minor radius of said vacuum vessel section; modular saddle coil means for generating a poloidal confining magnetic field, said saddle coil means disposed circumferentially around the major radius of said toroidal vacuum vessel section and encircling said toroidal field coil about the major radius of said toroidal field coil; and at least one modular ohmic heating coil, said ohmic heating coil including a magnetic coil having the same general shape and concentrically disposed around a toroidal field coil. 5. The tokamak module of claim 4 wherein said modular ohmic heating coil means comprises an ohmic heating coil having the same general shape as said toroidal field coil and disposed concentrically around said toroidal field coil. 6. The tokamak module of claim 5 wherein said ohmic heating coil is disposed around the outer perimeter of said toroidal field coil. 7. The tokamak module of claim 5 wherein said ohmic heating coil is wound around said toroidal field coil such that said ohmic heating coil encloses said toroidal field coil. 8. The tokamak module of claim 7 wherein said toroidal field coil is D-shaped. 9. The tokamak module of claim 8 wherein said saddle coil means comprises a first set of saddle coils operatively positioned to generate an equilibrium magnetic field and a second set of saddle coils operatively positioned to generate a divertor magnetic field. 10. A tokamak reactor system comprising: 11. The system of claim 10 wherein said toroidal field coil is D-shaped and said ohmic heating coil is disposed around the outer perimeter of a toroidal field coil. 12. The tokamak reactor system of claim 11 wherein said ohmic heating coil is horseshoe-shaped, the opening in said ohmic heating coil being aligned with the flat portion of said D-shaped toroidal field coil. 13. The tokamak reactor system of claim 11 comprising a plurality of said ohmic heating coils. 14. The tokamak reactor system of claim 1 wherein said modules include a plurality of toroidal field coils and a plurality of ohmic heating coils and wherein said ohmic heating coils are wound around said toroidal field coils, such that said ohmic heating coils enclose said toroidal field coils. 15. The tokamak reactor system of claim 14 wherein said toroidal coils are D-shaped. 16. The tokamak reactor system of claim 15 wherein said saddle coil means comprise a first set of saddle coils operatively positioned to generate an equilibrium magnetic field, and a second set of saddle coils operatively positioned to generate a divertor magnetic field. 17. The tokamak reactor system of claim 1 further comprising a trimming coil, said trimming coil comprising a conducting ring disposed around the outer circumference of said toroidal field coils |
claims | 1. A lithographic projection apparatus constructed and arranged to scan-image a mask pattern in a mask onto a substrate with a radiation sensitive layer, the mask having at least one transmissive region bounded at least in the scan direction by opaque regions, the lithographic projection apparatus comprising: a radiation system comprising a radiation source and an illumination system, said radiation system generating an illumination beam and controlling said illumination beam to scan said transmissive region between said opaque regions, said illumination beam comprising charged particles; a first movable object table provided with a mask holder constructed and arranged to hold a mask; a second movable object table provided with a substrate holder constructed and arranged to hold a substrate; and a projection system constructed and arranged to image an irradiated portion of the mask onto a target portion of the substrate; wherein said illumination system is embodied to begin a scan by generating a relatively narrow beam adjacent one of said opaque regions, to increase the width of said beam whilst it remains adjacent said opaque region, and to scan said beam towards the other of said opaque regions, thereby to provide a substantially uniform exposure on the substrate. 2. An apparatus according to claim 1 , wherein said illumination system is embodied at the end of the scan, when said beam approaches said other opaque region, to reduce the width of said beam whilst said beam is adjacent said other opaque region. claim 1 3. An apparatus according to claim 1 , wherein said illumination system is embodied substantially to avoid illuminating said opaque regions. claim 1 4. An apparatus according to claim 1 , wherein said illumination system is embodied such that the change n the width of said beam is performed substantially linearly at a rate equal to a speed at which said transmissive region is scanned. claim 1 5. An apparatus according to claim 1 , wherein said mask comprises a reticle divided into a plurality of elongate sub-fields separated by struts, said transmissive region comprising said sub-fields and said opaque regions comprising said struts. claim 1 6. An apparatus according to claim 5 , wherein said illumination system is embodied to scan along the length of said sub-fields and said apparatus further comprises a transverse scanning unit constructed and arranged to scan said mask and said substrate perpendicularly to said sub-fields. claim 5 7. An apparatus according to claim 6 , wherein said illumination system is embodied to scan the length of a sub-field a plurality of times between operations of said transverse scanning unit. claim 6 8. An apparatus according to claim 1 , wherein said illumination system comprises first and second shaping apertures and a first deflection unit constructed and arranged to displace said image of said first shaping aperture, thereby controlling the width of said illumination beam. claim 1 9. An apparatus according to claim 8 , wherein said illumination system further comprises a second deflection unit constructed and arranged to displace said illumination beam on said mask. claim 8 10. An apparatus according to claim 8 , wherein said illumination system further comprises a first focusing unit constructed and arranged to project an image of said first shaping aperture onto the plane of said second shaping aperture. claim 8 11. An apparatus according to claim 1 , wherein the projection system includes a projection focusing unit controlled in accordance with the width of or current in said projection beam. claim 1 12. An apparatus according to claim 1 , wherein said illumination beam has a hexagonal cross-section at the mask, and said illumination system is embodied to vary the width of said illumination beam parallel to said scan direction. claim 1 13. A method of manufacturing a device using a lithographic projection apparatus comprising: a radiation system comprising a radiation source and an illumination system, said radiation system generating an illumination beam, said illumination beam comprising charged particles; a first object table provided with a mask holder constructed and arranged to hold a mask; a second object table provided with a substrate holder constructed and arranged to hold a substrate; and a projection system constructed and arranged to image an irradiated portion of the mask onto a target portion of the substrate provided with a radiation sensitive layer; the method comprising: scan-imaging a mask pattern in said mask onto said target portion of said substrate, the having at least one transmissive region bounded at least in the scan direction by opaque regions, wherein said imaging comprises: at the beginning of a scan, generating a relatively narrow beam adjacent one of said opaque regions, increasing the width of said beam while it remains adjacent said opaque region, and scanning said beam towards the other of said opaque regions, thereby to provide a substantially uniform exposure on the substrate. 14. A device manufactured using a lithographic projection apparatus comprising: a radiation system comprising a radiation source and an illumination system, said radiation system generating an illumination beam comprising charged particles; a first object table provided with a mask holder constructed and arranged to hold a mask; a second object table provided with a substrate holder constructed and arranged to hold a substrate; and a projection system constructed and arranged to image an irradiated portion of the mask onto a target portion of the substrate provided with a radiation sensitive layer; the method comprising: scan-imaging a mask pattern in said mask onto said target portion of said substrate, the mask having at least one transmissive region bounded at least in a scan direction by opaque regions, wherein said imaging comprises at the beginning of a scan, generating a relatively narrow beam adjacent one of said opaque regions, increasing a width of said beam whilst it remains adjacent said opaque region, and scanning said beam towards the other of said opaque regions, thereby to provide a substantially uniform exposure on the substrate. |
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description | Reference will now be made in detail to the description of the invention as illustrated in as the drawings with like numerals indicating like parts throughout the several views. As described in detail hereinafter, the present invention provides systems and methods for storing exothermic material, such as spent nuclear fuel (SNF), among others. Although the present invention will be described herein in relation to the storage of SNF, it should be noted that applications of the teachings of the present invention are not so limited, with other such applications being considered well within the scope of the present invention. As depicted in FIG. 1, a preferred embodiment of the storage system 100 of the present invention incorporates an overpack 102 (shown schematically) and a canister assembly 104, which includes an inner canister 106 and an outer canister 108. Preferably, the inner canister is cylindrically shaped and provides an inner storage volume 110 which is defined, at least in part, by canister wall 111. Although depicted in FIG. 1 as being cylindrically shaped, the inner canister as well as the outer canister may be provided in various shapes, provided the canisters may appropriately receive material for storage. Preferably, the outer canister provides an additional storage volume 112, which is adapted to be oriented about at least a portion of the inner storage volume 110. Storage volume 112 is defined, at least in part, by inner and outer walls 114 and 116, respectively, and a bottom (not shown). So configured, exothermic material, such as SNF, for example, may be stored within either or both of the storage volumes 110 and 112. Referring now to FIG. 2, canister assembly 104 will be described in greater detail. In the embodiment depicted in FIG. 2, inner canister 106 is provided with a cylindrical exterior shape and outer canister 108 is provided in an annular configuration. As mentioned hereinbefore, however, various other configurations of inner and outer canisters may be utilized, with all such shapes and configurations being considered well within the scope of the present invention. It is preferred, however, that the inner canister be adapted to be received within a canister-receiving volume 130 of the outer or second canister while allowing a sufficient volume or clearance for a cooling medium flow between the canisters. In the embodiment depicted in FIG. 2, cooling medium flow between the canisters preferably is, at least partially, facilitated by one or more flow channels 140 which are provided between the first canister wall and the inner wall of the second canister. The outer wall of the second canister also may serve as a cooling surface over which cooling medium flow may be directed, e.g., an outer cooling medium flow channel(s) may be formed between the outer wall of the second canister and the overpack. In some embodiments, cooling medium flows over the various walls of the canisters may be facilitated by one or more flow orifices (e.g., orifices 141 and 143 of FIG. 1). Such flow orifices may be formed through various portions of the overpack, such as through the overpack lid and/or sidewalls. Additionally, a support structure or pedestal (not shown) may be provided which is adapted to maintain the canisters in a spaced relationship with the bottom or floor of the overpack, thereby allowing a cooling medium to flow beneath the canisters. For instance, in the embodiment depicted in FIG. 1, a cooling medium may enter the overpack through flow orifice 141, and may be directed toward the canisters by conduit 145. So configured, the cooling medium, such as air, water or other heat removal agents, may flow across and between the various walls of the canisters and/or of the overpack, thereby potentially significantly increasing the effective heat transfer area, such as by more than fifty percent (50%), over prior art canister designs. Flow channels 140 preferably are formed, at least in part, by spacers 142, which engage between the canisters and which maintain the canisters in a spaced configuration relative to each other, although various other configurations may be utilized. As depicted in the accompanying figures, one or more spacers may be suitably adapted to be received within an alignment channel 146 which, in addition to aiding in alignment of the inner canister within the canister-receiving volume, may prevent the inner canister from rotating about its longitudinal axis or, otherwise, jostling within the inner storage volume. It should be noted that spacers 142 are depicted as elongated components affixed to the inner wall of canister 108 and the alignment channels are depicted as elongated components affixed to the wall of canister 106; however, alternative configurations may be utilized. For example, the spacers may be affixed to the wall of canister 106 with the channels being formed on the inner wall of canister 108. As an additional example, the channels and spacers may be formed as less than full length segments engaging the various canisters. Referring once again to FIG. 2, the outer canister 108 will now be described in greater detail. Preferably, outer canister 108 is adapted to store fuel assemblies 150 in a prescribed pattern between its inner and outer walls. In the embodiment depicted in FIG. 2, the annular shape of the outer canister typically results in the formation of wedge-shaped spaces 152 between the various fuel assemblies. Depending upon the particular application, spaces 152 may be retained as voids between the fuel assemblies or may be, at least partially, filled by a material for facilitating neutron moderation and absorption, shielding, cooling, positioning and/or protecting of the fuel assemblies. For instance, when the storage system is adapted for storing spent nuclear fuel, one or more of the spaces 152 may be occupied by a material containing neutron absorbers. As described hereinbefore, the inner canister 106 is adapted to be received within a canister-receiving volume 130 of the outer canister 108. Maintaining the inner canister within the canister-receiving volume preferably is facilitated by the inner canister engaging a bottom structure of the outer canister. In the embodiment depicted in FIG. 2, such a bottom structure is provided in the form of an array of beams 160 (although various other configurations may be utilized) which are sufficiently durable so as to enable the inner canister to be supported and/or carried by the outer canister, such as during repositioning of the canisters, for instance. The array of beams configuration also provides the added benefit of allowing a cooling medium to flow upwardly through the beams and between the canisters, thereby promoting effective cooling of the storage system. Depending upon the particular application, either or both of the inner and outer canisters may be provided with suitable lids for sealing materials stored by the canisters therein. In some applications, however, the use of one or more lids may not be desirable. For instance, and not for the purpose of limitation, while storing materials that produce gasses, sealing of such materials in a lidded canister may provide less than adequate venting from the canister of the produced gasses, thereby potentially compromising the structural integrity of the canister due to excess gas pressure created within the canister. As described herein in relation to a preferred embodiment, storage system 100 potentially provides for high density storage of exothermic materials, e.g., SNF, while improving the heat transfer area typically provided by long-term dry storage applications. For example, extraction of one hundred percent (100%) to one hundred fifty percent (150%) or more heat from a given volume of canisterized fuel may be attained while maintaining the temperature of the material in and of the storage canisters at acceptable levels. Thus, the storage of very hot canisterized fuel may be accomplished without exceeding material, e.g., steel, concrete, neutron shielding, or SNF temperature limits. The foregoing description has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise forms disclosed. Obvious modifications or variations are possible in light of the above teachings. The embodiment or embodiments discussed, however, were chosen and described to provide the best illustration of the principles of the invention and its practical application to thereby enable one of ordinary skill in the art to utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. All such modifications and variations, are within the scope of the invention as determined by the appended claims when interpreted in accordance with the breadth to which they are fairly and legally entitled. |
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