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055531067	description	DESCRIPTION OF PREFERRED EMBODIMENTS First of all, the first principle of the present invention will be described below with reference to FIG. 2. The first principle resides in a method of impinging, against a member surface in water, a high speed water jet at temperature lower than that of the water or the member surface to produce thermal shock stress in a member used in a reactor pressure vessel, thereby reducing tensile residual stress of the member in the reactor pressure vessel or transforming it into compressive stress. The first principle will be described below in more detail with reference to an apparatus shown in FIG. 2. As shown in FIG. 2, high temperature water 2 of 40.degree. C. to 100.degree. C. is filled in a high temperature tank 3. A metallic member 1 is set to be submerged in the high temperature water 2. A nozzle 4 is set to be submerged in the high temperature water 2. Cold water 6 at low temperature (e.g., 20.degree. C.) is pumped under pressure by a high pressure pump 5 and is introduced to the nozzle 4 through a conductor 7. Therefore, a cold water jet 8 in the form of a high speed submerged water jet is ejected from the nozzle 4. The cold water jet 8 ejected from the nozzle 4 impinges against a surface of the metallic member 1. The nozzle 4 is moved in the horizontal direction (as indicated by arrow) in FIG. 2 along the surface of the metallic member 1 while ejecting the cold water jet 8. Next, the operation of improving residual stress will be described with reference to FIG. 3. FIG. 3 is a graph showing the stress (.epsilon.)-strain (.sigma.) curve of a metallic member shown 1. The vertical axis of the graph represents stress that is tensile stress in the upper half and compressive stress in the lower half, whereas the horizontal axis represents strain that is tensile strain in the right half and compressive strain in the left half Generally, tensile residual stress on the order of yield stress exists in the surface of the metallic member 1 as a result of manufacturing process, welding and so forth in an initial state. This initial state is indicated by a point (1) on the stress-strain curve shown in FIG. 3. At the beginning, a surface layer of the metallic member 1 has the same temperature as the high temperature water 2. In a certain region against which the cold water jet 8 impinges (hereinafter referred to as an impingement region), however, there occurs a transient temperature difference .DELTA.T between the surface layer and a subsurface layer of the metallic member 1 because of rapid cooling by the cold water jet 8. The temperature difference .DELTA.T between the surface layer and the subsurface layer brings about a thermal shock to cause fictitious tensile stress .DELTA..sigma., the actual tensile stress .DELTA..sigma..sub.act and tensile strain .DELTA..epsilon. in the surface layer. The fictitious tensile stress .DELTA..sigma. and the tensile strain .DELTA..epsilon. caused at this time are expressed below: EQU .DELTA..epsilon.=.alpha..times..DELTA.T.times.F EQU .DELTA..sigma.=E.times..DELTA..epsilon./(1-.nu.) where .alpha.: coefficient of linear thermal expansion PA1 .nu.: Poisson's ratio PA1 F: strain restriction Coefficient (F.revreaction.1 in this case) PA1 E: modulus of elasticity Supposing now a residual stress improving operation to be carried out in inspecting a reactor pressure vessel, for example, the resulting fictitious tensile stress and tensile strain are given, respectively, by .DELTA..sigma..revreaction.300 MPa and .DELTA..epsilon..revreaction.0.15% on condition of the coefficient of linear thermal expansion .alpha.=17.times.10.sup.-6 mm/mm.degree. C., Poisson's ratio .nu.=0.3, the strain restriction coefficient F=1, and the coefficient of longitudinal elasticity E=2.times.10.sup.5 MPa, taking into account the physical properties of austenite stainless steel or Inconel used as a material for members in the reactor pressure vessel, as well as .DELTA.=60.degree. C., taking into account a relatively high temperature of the reactor water during the inspection. If the initial residual stress due to the manufacturing process, etc. does not exist in the metallic member 1, the thermal shock stress of about 300 MPa would be wholly developed in the metallic member 1. But when tensile stress close to yield stress remains already in the metallic member 1, as mentioned above, the surface layer is partly yielded while strain of .DELTA..epsilon..revreaction.0.15% is developed and, therefore, the stress is not so increased in the surface layer. Thus, the surface state is changed from the point (1) to a point (4) in FIG. 3. After that, the impingement region of the metallic member 1 rapidly cooled by the impingement of the cold water jet 8 is heated again to the same temperature as the high temperature water 2, whereupon the transiently occurred temperature difference .DELTA.T disappears. Since the thermal shock stress resulted from the temperature difference also disappears, the impingement region in the surface layer of the metallic member 1 elastically reverts from the partly yielded state, under the elastic restriction. Accordingly, the surface state is changed from the point (4) to a point (5) in FIG. 3. In other words, based on the first principle described above, the residual stress improving method of this embodiment can improve the surface state of the metallic member from the initial tensile residual stress state of (1) to a compressive residual stress state of (5). Even when the temperature of the cold water jet 8 is not so low and the transient temperature difference .DELTA.T between the surface layer and the subsurface layer is relatively small, the surface state can be changed from the point (1).fwdarw.point (2).fwdarw.point (3), for example, in FIG. 3 depending on a value of the temperature difference. It is thus possible to improve the surface state of the metallic member from the initial tensile residual stress state of (1) to a state of (3) where the tensile stress is reduced. Thus, according to the first principle, the stress state of a metallic member is improved by impinging, against the metallic member, a cold water jet at temperature that is low enough to produce a temperature difference with respect to the ambient water temperature,i and causing thermal shock stress in a surface layer of the metallic member. Also, according to the first principle, since cavitation bubbles are not positively utilized, there is no necessity of using a pump of great horse power to increase a flow speed of the cold water jet ejected from the nozzle and, hence, the size of necessary equipment is held down. The second principle will be described below with reference to FIG. 4. Arrangements shown in FIG. 4 are different from those shown in FIG. 2 for explaining the first principle in that a nozzle 18 adapted to accelerate cavitation (e.g., a horn-shaped nozzle which ejects a water jet in the form broadening toward the end) is provided at the end of the conductor 7 instead of the normal nozzle 4, thereby ejecting a cold water cavitation jet 19 including cavitation bubbles. The remaining arrangements are the same as shown in FIG. 2, and identical components to those in FIG. 2 are denoted by the same reference numerals and will not be here described. With the residual stress improving method based on the second principle, as shown in FIG. 4, a water jet ejected from the nozzle 18 is provided as the cold water cavitation jet 19 including cavitation bubbles and, therefore, a high speed turbulent flow is generated to enhance the cooling effect. As a result, stress is caused by a stronger thermal shock and the residual stress improving effect is enhanced correspondingly. Further, when the cavitation bubbles impinge against the member surface and collapse there, high pressure is produced to induce bearing stress in the surface of the metallic member 1. Therefore, the residual stress improving effect by the so-called peening is developed in addition to the thermal effect of improving residual stress, which maximizes the residual stress improving effect. Thus, according to the second principle, since the residual stress improving effect by the peening utilizing cavitation bubbles is added to the residual stress improving effect by a thermal shock, even if the cavitation bubbles are not so sufficiently produced as compared with the case of achieving the residual stress improving effect by only the peening utilizing cavitation bubbles, the satisfactory residual stress improving effect can be achieved with the stress improvement additionally enhanced by applying a thermal shock. As a result, these is no necessity of using a pump of great horse power to increase a flow speed of the cold water jet ejected from the nozzle and, hence, the size of necessary equipment is held down. A first embodiment of the present invention based on the first principle will now be described below with reference to FIGS. 1, 5 and 6. This embodiment intends to improve residual stress of a core shroud as a member in a pressure vessel of a boiling water reactor based on the above-explained first principle. Prior to starting the method of improving residual stress of the member in a reactor pressure vessel according to this embodiment, a top cover of a pressure vessel 9 of a reactor 21 is removed, and a steam drier, a steam separator and fuel assemblies (not shown) are taken out successively. Then, the pressure vessel 9 is filled with reactor water 11 at high temperatures (40.degree. C. to 100.degree. C.) completely to a level above a top guide 10. That condition is illustrated in FIG. 5. The temperature of the reactor water 11 is adjustable with friction heat by driving a recirculation pump associated with the nuclear plant or reactor. Next, a nozzle 14 connected through a conductor 17 to a low temperature water tank 12 and a high pressure pump 15, which are installed outside the reactor 21, is introduced into the pressure vessel 9 by a driving mechanism (not shown) such that the nozzle is moved to face a predetermined position of the core shroud 13. The driving mechanism may be of the same arrangements as those of a water jet peening apparatus disclosed in JP, A, 5-78738, for example. More specifically, the driving mechanism comprises a circumferentially movable carriage provided at a top of the pressure vessel 9 to be able to move in the circumferential direction of the pressure vessel 9, a radially movable carriage provided on an upper surface of the circumferentially movable carriage to be able to move in the radial direction of the pressure vessel 9, and a mast suspended from the radially movable carriage and divided into multiple stages to be able to telescopically extend and contract in the vertical direction. The nozzle 14 is attached to a lowest end of the mast 14. Though not specifically shown, the conductor 17 has a divided structure similarly to the mast of the driving mechanism so that it can also telescopically extend and contract in the vertical direction or in the radial direction. With such arrangements of the driving mechanism and the conductor 17, the nozzle 14 can be radially and vertically moved in the pressure vessel 9 and smoothly guided to the predetermined position in the pressure vessel 9 while the conductor 17 is kept connected to the high pressure pump 15. Subsequently, cold water (e.g., 20.degree. C.) in the low temperature water tank 12 is pumped by the high pressure pump 15 and is supplied to the conductor 17 under pressure. The supplied water is ejected as a high speed cold water jet 8 from the nozzle 14 to impinge against a predetermined region of the core shroud 13 in the reactor water 11 at high temperature. That condition is illustrated in FIG. 1. At this time, an ejection speed of the cold water jet 8 is preferably not less than 100 m/s in order to increase the heat conductivity between the cold water jet 8 and a surface layer of the core shroud 13 and to achieve the satisfactory residual stress improving effect, but the ejection speed is also preferably not larger than 700 m/s from limitations on a capability of the high pressure pump 15. Further, taking into account that scales are peeled off from the core shroud 13 and may give rise to an adverse effect due to their grinding action if the cold water jet 8 exceeds a certain high speed, and that a sufficient allowance must be provided in reliability and durability of the equipment, such as the high pressure hose and the high pressure pump, in view of a great reaction force caused by the ejection of the cold water jet and severe radiation environment, the ejection speed of the cold water jet 8 is more preferably in the range of 200 m/s to 400 m/s. Additionally, in consideration of operability of the nozzle and handling facility of the high pressure hose, the ejection speed of the cold water jet 8 is most preferably in the range of 250 m/s to 350 m/s. After that, while continuing to eject the cold water jet 8 from the nozzle 14, the nozzle 14 is vertically moved by the above-mentioned driving mechanism (not shown) so that the nozzle is reciprocated to repeatedly impinge the cold water jet 8 against the same region several times. FIG. 6 illustrates the condition where the nozzle 14 is moved to a lower position. The nozzle 14 may be moved radially or circumferentially rather than vertically, or may be reciprocated by combined movements in the vertical direction and the radial or circumferential direction. With this embodiment, in a first step, since the cold water jet 8 is ejected to impinge against and cool a surface layer of the core shroud 13 in a tensile stress residual state, the surface layer is subjected to tensile stress by a thermal shock to cause tensile strain and, simultaneously, it is partly yielded. In a second step, since the nozzle 14 is moved away from the impingement region, the partly yielded surface layer becomes free from the cold water jet 8 and is heated again by the reactor water 11 at high temperature. Accordingly, since the thermal shock stress applied to the surface layer disappears, the surface layer elastically reverts from the partly yielded state, for improvement to a state where tensile stress is relieved or to a compressive residual stress state. It is thus possible to prevent the occurrence of intergranular stress corrosion cracking. Also, the residual stress improving operation can be simply performed by simple arrangements with no need of disassembling or dismantling the core shroud. When oxide scales or other contaminants are deposited on the surface layer of the core shroud 13, the oxide scales or the other contaminants are contracted by rapid cooling with the cold water jet 8 to cause shear strain at the boundary between the contaminants and the surface layer, allowing the contaminants to be easily peeled off. In other words, cleaning of the surface contaminants on the core shroud 13 can be accelerated. It is thus possible to prevent crevice aided stress corrosion cracking, fatigue cracking or the like induced by the oxide scales or the other contaminants. Further, since the cold water jet 8 is ejected so as to repeatedly impinge against the same region by reciprocating the nozzle 14, the cleaning action is further increased. While the core shroud 13 is cited in the foregoing embodiment as one example of the members of in the reactor pressure vessel to which the residual stress improving method is applied, the present invention is not limited to the illustrated embodiment, but also applicable to other reactor equipment, wall surfaces and so on. While the temperature of the cold water jet 8 ejected from the nozzle 14 is set to 20.degree. C. in the foregoing embodiment, the present invention is not limited to the illustrated embodiment. So long as the temperature of the cold water jet 8 is lower than any temperatures of the reactor water 11 and members in the reactor pressure vessel to be treated, the present invention can provide the similar effect. However, the more satisfactory residual stress improving effect by thermal shock stress is expected with the larger temperature difference. Therefore, the temperature of the cold water jet 8 is preferably held in the range of 0.degree. C. to 40.degree. C. Further, while the foregoing embodiment is arranged so as to move the nozzle 14 while ejecting the cold water jet 8 from the nozzle 14 so that the surface layer is rapidly cooled by the cold water jet 8 and then heated again by the reactor water 11, the present invention is not limited to the illustrated embodiment. By way of example, a valve means or the like capable of opening and closing may be attached to the conductor 17, and the ejection of the cold water jet 8 from the nozzle 14 may be stopped by closing the valve means with the nozzle 14 kept intact. In this modified case, the present invention can also provide the similar effect. This method is particularly suitable for the case where the nozzle 14 is provided in plural number to perform the residual stress improving operation at a time or in a short time over a wide area of the member in the reactor pressure vessel to be treated. While the normal nozzle 14 is attached to the end of the conductor 17 in the foregoing embodiment, the nozzle 18 adapted to accelerate cavitation may be used instead for utilization of the second principle explained in connection with FIG. 4. In this case, as described above, not only the residual stress improving effect, but also the effect of cleaning surface contaminants can be achieved at maximum. A second embodiment of the present invention will be described below with reference to FIG. 7. This embodiment intends to use pure water as being ejected to form the cold water jet. FIG. 7 shows arrangements for ejecting the cold water jet according to a residual stress improving method of this embodiment. Note that identical components in FIG. 7 to those in the first embodiment are denoted by the same reference numerals. One point of differences between the arrangements of FIG. 7 and those of the first embodiment shown in FIG. 1 is that a low temperature pure water tank 28 is provided instead of the low temperature tank 12 to use pure water as working water. Specifically, the pure water in the low temperature pure water tank 28 is pumped by the high pressure pump 15 and is ejected under pressure as a cold pure water jet 16 from the nozzle 14 through the conductor 17. Another point of differences in arrangements between FIG. 7 and FIG. 1 is that a suction hose 27 is provided to suck or collect oxide scales or other contaminants peeled off. The suction hose 27 is introduced into the pressure vessel 9 by a driving mechanism (not shown) similar to that used for the nozzle 14, or the driving mechanism for the nozzle 14, and is movable in the vertical and radial directions as with the nozzle 14. The suction hose 27 is connected to a contaminant treating apparatus (not shown) installed outside the pressure vessel 9, and the oxide scales or the like sucked through the suction hose 27 is conveyed to the contaminant treating apparatus for treatment. The oxide scales or the like which have not been sucked through the suction hose 27 are treated by a cleaning apparatus (not shown) associated with the reactor 21 as the reactor water is recirculated. The other arrangements and the operation steps are substantially the same as in the first embodiment. With this embodiment, since the cold pure water jet 16 is ejected from the nozzle 14 by using the pure water as working water, the reactor water 11 will not be contaminated by addition of the working water. It is thus possible to minimize contamination of the reactor water. While the low temperature pure water tank 28 is provided in the foregoing embodiment, the present invention may also be practiced by branching a pipe from a pure water storing tank installed in the nuclear plant for resupply of the reactor water or other purposes and supplying pure water through the pipe, without providing the low temperature pure water tank 28. This modification is effective to eliminate the need of installing a separate tank. Since the temperature of the pure water in the pure water storing tank is usually about 20.degree. C., the stored pure water can be suitably used as the working water to form the cold pure water jet 16. While water outside the reactor is used as the working water in the second embodiment, the reactor water may be used instead. The case of using the reactor water will be described below as a third embodiment With reference to FIG. 8. Note that identical components in FIG. 8 to those in FIG. 7 are denoted by the same reference numerals. Arrangements of this embodiment shown in FIG. 8 are different from those of the Second embodiment shown in FIG. 7 in that the suction hose 27 is connected through a suction pump 31 to a cleaning apparatus 32 for separating contaminants and pure water, the cleaning apparatus 32 is connected to a cooler 33, and the cooler 33 is connected to a cold reactor water tank 29. Specifically, the reactor water sucked through the suction hose 27 conveyed to the cleaning apparatus 32 for separation from contaminants. The cleaned water is cooled by the cooler 33 and then supplied to the cold reactor water tank 29. The reactor water in the cold reactor water tank 29 is pumped by the high pressure pump 15 and is ejected under pressure as a high speed cold reactor water jet 30 from the nozzle 14 through the conductor 17. Incidentally, the contaminants separated by the cleaning apparatus 32 is sent to a contaminant treating apparatus (not shown) for treatment. With this embodiment, the reactor water 11 will not be contaminated by addition of the working water into the reactor pressure vessel as with the second embodiment. Also, since the cold reactor water jet 30 and the water supplied to the high pressure pump 15 are recirculated through a closed loop, the reactor water 11 is held at a substantially constant level in the reactor. Therefore, adjustment of the water level in the reactor is not almost required or much simplified when the residual stress improving operation is completed. Further, since the cooler 33 is provided, the temperature difference .DELTA.T can be enlarged and the greater residual stress improving effect can be achieved correspondingly. A fourth embodiment of the present invention will be described below with reference to FIG. 9. In this embodiment, an elbow-shaped nozzle is used as the nozzle. FIG. 9 shows arrangements for ejecting the cold water jet according to a residual stress improving method of this embodiment. Note that identical components in FIG. 9 to those in the first and second embodiments are denoted by the same reference numerals. The arrangements shown in FIG. 9 are different from those of the first embodiment in that an elbow-shaped nozzle 24 for ejecting a high speed cold water jet 25 at an angle .theta. with respect to the direction in which the cold water is introduced through the conductor 17 (i.e., the downward direction indicated by arrow in FIG. 9), enabling the desired operation to be performed in a narrow space 26 inside an illustrated cylindrical member 34, for example. The other arrangements except the nozzle and the operation steps are substantially the same as in the first embodiment. With this embodiment, since the elbow-shaped nozzle 24 is used as the nozzle, the residual stress improving operation can be performed not only in a wide space, but also in the narrow space 26 by ejecting the cold water jet 25 to impinge against the member surface in the reactor pressure vessel. It is thus possible to cause stress by a high thermal shock and to enhance both the residual stress improving effect and the cleaning effect. Although the angle .theta. is selected from the range of 180.degree. to 90.degree. depending on narrowness of the space 26 in which the operation is to be carried, it is preferably set to 90.degree. for the purpose of achieving the residual stress improving effect and the cleaning effect as high as possible. The following features can be provided by any of the foregoing embodiments. As the first step, the water jet in the form of a high speed submerged water jet at low temperature is ejected to impinge against and cool a partial surface region of the member in the reactor pressure vessel, the surface region being in a tensile stress residual state. Therefore, the surface layer is subjected to tensile stress by a thermal shock to cause tensile strain and, simultaneously, it is partly yielded. Then, as the second step, the impingement of the water jet against the partial surface region of the member in the reactor pressure vessel is stopped, i.e., cooling of the partial surface region, is finished, allowing the same region to be heated against to the same level as the temperature of the surrounding water environment. Therefore, the thermal shock stress disappears, and the partial surface region of the member in the reactor pressure vessel elastically reverts from the partly yielded state, for improvement to a state where tensile stress is relieved or to a compressive residual stress state. Accordingly, the occurrence of intergranular stress corrosion cracking can be prevented. When oxide scales or other contaminants are deposited on the surface layer, the oxide scales or the other contaminants are contracted by rapid cooling to cause shear strain at the boundary between the contaminants and the surface layer, allowing the contaminants to be easily peeled off. Therefore, cleaning of the contaminants on the members in reactor pressure vessel surface can be accelerated. As a result, crevice aided stress corrosion cracking, fatigue cracking or the like induced by the oxide scales or the like can be prevented. Further, the residual stress improving operation can be simply performed by simple arrangements with no need of disassembling or dismantling the core shroud in the reactor pressure vessel. As will be seen from the foregoing embodiment, the present invention includes total eleven inventions from claim 1 (first invention) to claim 11 (eleventh invention) as defined in the attached claims. According to the first invention, in the first step, since the water jet in the form of a high speed submerged water jet at low temperature is ejected to impinge against and cool a first region of a surface of members in reactor pressure vessel in a tensile stress residual state, the temperature of only a surface layer in the first region against which the water jet impinges is lowered to produce the transient temperature difference .DELTA.T between the surface layer and a subsurface layer. Therefore, the surface layer is subjected to tensile stress by a thermal shock to cause tensile strain and, simultaneously, it is partly yielded. Then, in the second step, since the impingement of the water jet against the first region is stopped, i.e., cooling of the first region, is finished so that the first region is heated again, the temperature difference .DELTA.T is eliminated and the thermal shock stress disappears. Therefore, the surface layer in the first region of the surface of members in reactor pressure vessel reverts from the partly yielded state, for improvement to a state where tensile stress is reduced or to a compressive residual stress state. Accordingly, the occurrence of intergranular stress corrosion cracking can be prevented. When oxide scales or other contaminants are deposited on the surface layer, the oxide scales or the other contaminants are contracted by rapid cooling to cause shear strain at the boundary between the contaminants and the surface layer, allowing the contaminants to be easily peeled off. Therefore, cleaning of the contaminants on the surface of members in reactor pressure vessel can be accelerated. As a result, crevice aided stress corrosion cracking, fatigue cracking or the like induced by the oxide scales or the like can be prevented. Further, the residual stress improving operation can be simply performed by simple arrangements with no need of disassembling or dismantling structural members in a reactor pressure vessel. According to the second invention, in the second step, the nozzle is moved while ejecting the water jet from the nozzle such that the water jet impinges against the surface of members in reactor pressure vessel successively from the first region to a different second region. In addition to the operating advantages of the first invention, therefore, the second step of heating the first region again can be realized without stopping the ejection, of the water jet from the nozzle. According to the third invention, in the second step, the ejection of the water jet from the nozzle is stopped. In addition to the operating advantages of the first invention, therefore, the second step of heating the first region again can be realized without moving the nozzle. According to the fourth invention, an initial ejection speed of the water jet from the nozzle is not less than 100 m/s, but not larger than 700 m/s. In addition to the operating advantages of any of the first through third inventions, therefore, heat conductivity in the cooling process is increased to enhance the rapid cooling effect, and the satisfactory residual stress improving effect can be provided. According to the fifth invention, the initial ejection speed of the water jet from the nozzle is not less than 200 m/s, but not larger than 400 m/s. In addition to the operating advantages of any of the first through third inventions, therefore, a possibility that scales peeled off from the members in reactor pressure vessel materials may give rise to an adverse effect due to their grinding action, and a sufficient degree of reliability and durability of the equipment, such as the high pressure hose and the high pressure pump, can be ensured in view of even influences of a great reaction force caused by the ejection of the water jet and severe radiation environment. According to the sixth invention, the initial ejection speed of the water jet from the nozzle is not less than 250 m/s, but not larger than 350 m/s. In addition to the operating advantages of any of the first through third inventions, therefore, operability of the nozzle and handling facility of the high pressure hose are improved. According to the seventh invention, a source of the water jet is low temperature water obtained by cooling the reactor water and pumping the same under pressure. In addition to the operating advantages of any of the first through third inventions, therefore, the reactor water is held at a substantially constant level during the residual stress improving operation and, hence, adjustment of the water level in the reactor is not almost required or much simplified when the residual stress improving operation is completed. According to the eighth invention, the source of the water jet is low temperature water prepared outside the reactor and pumped under pressure. In addition to the operating advantages of any of the first through third inventions, therefore, a means for supplying the low temperature water to eject the water jet can be realized with no need of the specific cooler. According to the ninth invention, the source of the water jet is low temperature pure water prepared outside the reactor and pumped under pressure. In addition to the operating advantages of any of the first through third inventions, therefore, the reactor water at high temperature will not be contaminated by addition of working water. As a result, the means for supplying the low temperature water to eject the water jet can be realized while minimizing contamination of the reactor water. Also, if the pure water is obtained from a pure water storing tank installed in the nuclear plant for resupply of the reactor water or other purposes, there is no need of providing any separate low temperature water supplying means. According to the tenth invention, the water jet is high speed jet water including cavitation bubbles. In addition to the operating advantages of any of the first through third inventions, therefore, a high speed turbulent flow is generated to enhance the cooling effect, resulting in that stress is caused by a stronger thermal shock and the residual stress improving effect is enhanced correspondingly. Further, when the cavitation bubbles impinge against the surface of members in reactor pressure vessel material surface and collapse there, high pressure is produced to induce bearing stress in the surface of a member in the reactor pressure vessel. As a result, the residual stress improving effect by peening is developed in addition to the thermal effect of improving residual stress. According to the eleventh invention, the nozzle is a an elbow-shaped nozzle for ejecting the high speed submerged water jet at a predetermined angle with respect to the inflow direction of high pressure water supplied to the nozzle. In addition to the operating advantages of any of the first through third inventions, therefore, the water jet can be ejected to impinge against the surface of members in reactor pressure vessel even when the operation is to be carried out in a narrow space, enabling stress to be caused by a strong thermal shock.
051606968	claims	1. An apparatus for producing power from fertile nuclear materials and transmuting wastes therefrom to less radioactive species, said apparatus comprising in combination: a. means for generating a high intensity, high-energy beam of protons; b. a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons; c. a substantially gas-tight enclosure surrounding said spallation target; d. windowless means for directing the beam of protons onto the open surface of said spallation target; e. neutron moderation means for thermalizing neutrons generated from said spallation target; f. first means for containing the fertile nuclear material disposed within said neutron moderation means and spaced apart from and outside of said spallation target; g. second means for containing materials to be transmuted disposed within said neutron moderation means and spaced apart and outside of said spallation target, yet closer thereto than said first containment means; h. first flowing means for passing the fertile nuclear material and transmutation products thereof through said first containment means; i. means for combining the fertile nuclear material with a molten salt eutectic and causing the combination formed thereby to flow through said first flowing means; and j. means for extracting fission products from the molten salt eutectic flowing through said first flowing means, separating the stable and short-lived fission products therefrom, and introducing the remaining material into said second containment means for transmutation. a. means for generating a high intensity, high-energy beam of protons; b. a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons; c. a substantially gas-tight enclosure surrounding said spallation target; d. windowless means for directing the beam of protons onto the open surface of said spallation target; e. neutron moderation means for thermalizing neutrons generated from said spallation target; f. first means for containing the fissile nuclear materials disposed within said neutron moderation means and spaced apart from and outside of said spallation target; g. second means for containing material to be transmuted disposed within said neutron moderation means and spaced apart and outside of said spallation target, yet closer thereto than said first containment means; h. first flowing means for passing the fissile nuclear material and transmutation products thereof through said first containment means; i. means for combining the fissile nuclear material with a molten salt eutectic and causing the combination formed thereby to flow through said first flowing means; and j. means for extracting fission products from the molten salt eutectic flowing through said first flowing means, separating the stable and short-lived fission products therefrom, and introducing the remaining material into said second containment means for transmutation. a. means for generating a high intensity, high-energy beam of protons; b. a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons; c. a substantially gas-tight enclosure surrounding said spallation target; d. windowless means for directing the beam of protons onto the open surface of said spallation target; e. neutron moderation means for thermalizing neutrons generated from said spallation target; and f. first means for containing the material to be transmuted disposed within said neutron moderation means and spaced apart from and outside of said spallation target; g. second means for containing material to be transmuted disposed within said neutron moderation means and spaced apart and outside of said spallation target, yet closer thereto than said first containment means; h. first flowing means for passing the material to be transmuted and transmutation products thereof through said first containment means; i. means for combining the material to be transmuted with a molten salt eutectic and causing the combination formed thereby to flow through said first flowing means; and j. means for extracting transmutation products from the molten salt eutectic flowing through said first flowing means, separating the stable and short-lived fission products therefrom, and introducing the remaining material into said second containment means for transmutation. a. means for generating a high intensity, high-energy beam of protons; b. a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons; c. a substantially gas-tight enclosure surrounding said spallation target; d. windowless means for directing the beam of protons onto the open surface of said spallation target; e. neutron moderation means for thermalizing neutrons generated from said spallation target; f. first means for containing the higher actinide materials disposed within said neutron moderation means and spaced apart from and outside of said spallation target; and g. means disposed within said neutron moderation means and spaced apart from said spallation target for holding materials which generate tritium upon interaction with neutrons; h. second means for containing material to be transmuted disposed within said neutron moderation means and spaced apart and outside of said spallation target, yet closer thereto than said first containment means; i. first flowing means for passing the higher actinide materials to be transmuted and transmutation products thereof through said first containment means; j. means for combining the higher actinide materials to be transmuted with a molten salt eutectic and causing the combination formed thereby to flow through said first flowing means; and k. means for extracting transmutation products from the molten salt eutectic flowing through said first flowing means, separating the stable and short-lived fission products therefrom, and introducing the remaining material into said second containment means for transmutation. 2. The apparatus as described in claim 1, wherein the fertile nuclear materials are selected from the group consisting of .sup.238 U, .sup.232 Th, and mixtures thereof. 3. The apparatus as described in claim 1, wherein said spallation target includes high-Z material for production of neutrons by interaction with the high-energy beam of protons. 4. The apparatus as described in claim 3, wherein said liquid-metal spallation target includes a lead-bismuth eutectic mixture. 5. The apparatus as described in claim 4, further comprising first heat exchanger means through which the liquefied lead-bismuth eutectic is circulated in order to remove generated heat. 6. The apparatus as described in claim 1, wherein said high intensity, high-energy proton beam generation means provides protons having energies between 400 MeV and 10 GeV with an average proton beam current of greater than 10 ma. 7. The apparatus as described in claim 1, wherein said neutron moderation means includes heavy water. 8. The apparatus as described in claim 1, wherein said high intensity, high-energy proton beam generation means, said spallation target, and said neutron moderation means produce a thermal neutron flux sufficient to permit substantial two-neutron transmutation processes to occur in waste products resulting from producing power from the fertile nuclear material. 9. The apparatus as described in claim 1, further comprising second heat exchanger means for removing heat from the flowing molten salt eutectic and fertile nuclear material combination after the combination passes through said neutron moderation means. 10. The apparatus as described in claim 9, further comprising power generation means for generating electricity from heat removed by either or both of said first heat exchanger means and said second heat exchanger means, and for returning a portion of the electricity to said high intensity, high-energy proton beam generation means. 11. The apparatus as described in claim 1, wherein said windowless means for directing the high intensity, high-energy proton beam into said spallation target means includes a cooled, evacuated beam transport tube having one end thereof forming a substantially gas-tight seal to said gas-tight enclosure and disposed substantially vertically above said spallation target, the other end thereof forming a substantially gas-tight seal with said proton beam generating means; whereby volatile gases produced within said spallation target means may be removed from the vicinity of the high intensity, high-energy proton beam, and whereby the high intensity, high-energy proton beam may be directed substantially vertically and directly onto the open surface of said liquid-metal spallation target. 12. The apparatus as described in claim 5, wherein said gas-tight enclosure further comprises a convection air-cooled holding tank located below said first heat exchanger means into which the liquid-metal may drain, and normally-closed valve means responsive to a chosen operating condition demanding that said valve means be opened, thereby permitting the liquid metal to drain out of said first heat exchanger means and out of the region of said spallation target. 13. The apparatus as described in claim 1, further comprising means for sensing and maintaining the ion/fluoride valence balance in the molten salt eutectic. 14. The apparatus as described in claim 1, further comprising means for continuously removing .sup.233 Pa from the molten salt eutectic flowing through said flowing means in the event .sup.232 Th is utilized as a fertile material. 15. The apparatus as described in claim 1, further comprising second flowing means for passing the fission products through said second containment means. 16. An apparatus for producing power from fissile nuclear materials without necessity for long-term nuclear waste management, said apparatus comprising in combination: 17. The apparatus as described in claim 16, wherein the fissile nuclear materials are selected from the group consisting of .sup.235 U, .sup.239 Pu, and mixtures thereof. 18. The apparatus as described in claim 16, wherein said spallation target includes high-Z material for production of neutrons by interation with the high-energy beam of protons. 19. The apparatus as described in claim 18, wherein said liquid-metal spallation target includes a lead-bismuth eutectic mixture. 20. The apparatus as described in claim 19, further comprising first heat exchanger means through which the liquefied lead-bismuth eutectic is circulated in order to remove generated heat. 21. The apparatus as described in claim 16, wherein said high intensity, high-energy proton beam generation means provides protons having energies between 400 MeV and 10 GeV with an average proton beam current of greater than 10 ma. 22. The apparatus as described in claim 16, wherein said neutron moderation means includes heavy water. 23. The apparatus as described in claim 16, wherein said high intensity, high-energy proton beam generation means, said spallation target, and said neutron moderation means produce a thermal neutron flux sufficient to permit substantial two-neutron transmutation processes to occur in waste products resulting from producing power from the fissile nuclear material. 24. The apparatus as described in claim 16, further comprising second heat exchanger means for removing heat from the flowing molten salt eutectic and fissile nuclear material combination after the combination passes through said neutron moderation means. 25. The apparatus as described in claim 24, further comprising power generation means for generating electricity from heat removed by either or both of said first heat exchanger means and said second heat exchanger means, and for returning a portion of the electricity to said high intensity, high-energy proton beam generation means. 26. The apparatus as described in claim 16, wherein said windowless means for directing the high intensity, high-energy proton beam into said spallation target means includes a cooled, evacuated beam transport tube having one end thereof forming a substantially gas-tight seal to said gas-tight enclosure and disposed substantially vertically above said spallation target, the other end thereof forming a substantially gas-tight seal with said proton beam generating means; whereby volatile gases produced within said spallation target means may be removed from the vicinity of the high intensity, high-energy proton beam, and whereby the high intensity, high-energy proton beam may be directed substantially vertically and directly onto the open surface of said liquid-metal spallation target. 27. The apparatus as described in claim 20, wherein said gas-tight enclosure further comprises a convection air-cooled holding tank located below said first heat exchanger means into which the liquid-metal may drain, and normally-closed valve means responsive to a chosen operating condition demanding that said valve means be opened, thereby permitting the liquid metal to drain out of said first heat exchanger means and out of the region of said spallation target. 28. The apparatus as described in claim 16, further comprising means for sensing and maintaining the ion/fluoride valence balance in the molten salt eutectic. 29. The apparatus as described in claim 16, wherein said first fissile nuclear material containment means contains a sub-critical inventory of fissile material. 30. The apparatus as described in claim 1, further comprising second flowing means for passing the fission products through said second containment means. 31. An apparatus for transmuting higher actinide waste along with .sup.99 Tc, .sup.129 I, and other fission product waste, thereby eliminating necessity for long-term nuclear waste storage, said apparatus comprising in combination: 32. The apparatus as described in claim 31, wherein the higher actinide materials are selected from the group consisting of .sup.237 Np, .sup.241 Am, .sup.244 Cm, and mixtures thereof. 33. The apparatus as described in claim 31, wherein said spallation target includes high-Z material for production of neutrons by interaction with the high-energy beam of protons. 34. The apparatus as described in claim 33, wherein said liquid-metal spallation target includes a lead-bismuth eutectic mixture. 35. The apparatus as described in claim 34, further comprising first heat exchanger means through which the liquefied lead-bismuth eutectic is circulated in order to remove generated heat. 36. The apparatus as described in claim 31, wherein said high intensity, high-energy proton beam generation means provides protons having energies between 400 MeV and 10 GeV with an average proton beam current of greater than 10 ma. 37. The apparatus as described in claim 31, wherein said neutron moderation means includes heavy water. 38. The apparatus as described in claim 31, wherein said high intensity, high-energy proton beam generation means, said spallation target, and said neutron moderation means produce a thermal neutron flux sufficient to permit substantial two-neutron transmutation processes to occur in the material to be transmuted. 39. The apparatus as described in claim 31, further comprising second heat exchanger means for removing heat from the flowing molten salt eutectic and material to be transmuted combination after the combination passes through said neutron moderation means. 40. The apparatus as described in claim 39, further comprising power generation means for generating electricity from heat removed by either or both of said first heat exchanger means and said second heat exchanger means, and for returning a portion of the electricity to said high intensity, high-energy proton beam generation means. 41. The apparatus as described in claim 31, wherein said windowless means for directing the high intensity, high-energy proton beam into said spallation target means includes a cooled, evacuated beam transport tube having one end thereof forming a substantially gas-tight seal to said gas-tight enclosure and disposed substantially vertically above said spallation target, the other end thereof forming a substantially gas-tight seal with said proton beam generating means; whereby volatile gases produced within said spallation target means may be removed from the vicinity of the high intensity, high-energy proton beam, and whereby the high intensity, high-energy proton beam may be directed substantially vertically and directly onto the open surface of said liquid-metal spallation target. 42. The apparatus as described in claim 35, wherein said gas-tight enclosure further comprises a convection air-cooled holding tank located below said first heat exchanger means into which the liquid-metal may drain, and normally-closed valve means responsive to a chosen operating condition demanding that said valve means be opened, thereby permitting the liquid-metal to drain out of said first heat exchanger means and out of the region of said spallation target. 43. The apparatus as described in claim 31, further comprising means for sensing and maintaining the ion/fluoride valence balance in the molten salt eutectic. 44. The apparatus as described in claim 31, wherein a chosen quantity of .sup.239 Pu is added to the molten salt eutectic to generate sufficient heat to power said proton beam generating means, and to provide additional neutrons, while maintaining sub-criticality. 45. The apparatus as described in claim 31, further comprising second flowing means for passing the fission products through said second containment means. 46. An apparatus for simultaneously transmuting higher actinide materials and producing tritium without necessity for long-term nuclear waste storage, said apparatus comprising in combination: 47. The apparatus as described in claim 46, wherein the higher actinide materials are selected from the group consisting of .sup.237 Np, .sup.241 Am, .sup.244 Cm, and mixtures thereof. 48. The apparatus as describe in claim 46, wherein the materials which generate tritium upon interaction with neutrons are selected from the group consisting of .sup.3 He, .sup.6 Li, and mixtures thereof. 49. The apparatus as described in claim 46, wherein said spallation target includes high-Z material for production of neutrons by interation with the high-energy beam of protons. 50. The apparatus as described in claim 49, wherein said liquid-metal spallation target includes a lead-bismuth eutectic mixture. 51. The apparatus as described in claim 50, further comprising first heat exchanger means through which the liquefied lead-bismuth eutectic is circulated in order to remove generated heat. 52. The apparatus as described in claim 46, wherein said high intensity, high-energy proton beam generation means provides protons having energies between 400 MeV and 10 GeV with an average proton beam current of greater than 10 ma. 53. The apparatus as described in claim 46, wherein said neutron moderation means includes heavy water. 54. The apparatus as described in claim 46, wherein said high intensity, high-energy proton beam generation means, said spallation target, and said neutron moderation means produce a thermal neutron flux sufficient to permit substantial two-neutron transmutation processes to occur in the higher actinide materials. 55. The apparatus as described in claim 46, further comprising second heat exchanger means for removing heat from the flowing molten salt eutectic and higher actinide materials combination after the combination passes through said neutron moderation means. 56. The apparatus as described in claim 55, further comprising power generation means for generating electricity from heat removed by either or both of said first heat exchanger means and said second heat exchanger means, and for returning a portion of the electricity to said high intensity, high-energy proton beam generation means. 57. The apparatus as described in claim 46, wherein said windowless means for directing the high intensity, high-energy proton beam into said spallation target means includes a cooled, evacuated beam transport tube having one end thereof forming a substantially gas-tight seal to said gas-tight enclosure and disposed substantially vertically above said spallation target, the other end thereof forming a substantially gas-tight seal with said proton beam generating means; whereby volatile gases produced within said spallation target means may be removed from the vicinity of the high intensity, high-energy proton beam, and whereby the high intensity, high-energy proton beam may be directed substantially vertically and directly onto the open surface of said liquid-metal spallation target. 58. The apparatus as described in claim 51, wherein said gas-tight enclosure further comprises a convection air-cooled holding tank located below said first heat exchanger means into which the liquid-metal may drain, and normally-closed valve means responsive to a chosen operating condition demanding that said valve means be opened, thereby permitting the liquid metal to drain out of said first heat exchanger means and out of the region of said spallation target. 59. The apparatus as described in claim 46, further comprising means for sensing and maintaining the ion/fluoride valence balance in the molten salt eutectic. 60. The apparatus as described in claim 46, wherein sufficient .sup.239 Pu is added to the molten salt eutectic to generate sufficient heat to power said proton beam generating means, and to provide additional neutrons.
050948040	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS In describing the preferred embodiments of the invention, reference is first made to FIG. 1 in order to describe the preferred method for making high temperature nuclear fuel elements. Next, the other figures of the drawing are referred to in connection with the disclosure of several preferred high temperature nuclear fuel element structures that are made according to the invention. Nuclear fuel elements made in accordance with the present invention are effective to localize molten fissionable material within the pores of carbon or graphite bodies. In these elements the localized fissionable material is shielded from contact with high temperature moderating gases by providing a coating of diamond or other suitable material over the pores within which the fissionable material is localized. It is recognized that prior art nuclear materials intended for use in gas moderated reactors which characteristically have operating temperatures around 1000.degree. C. or less, have been developed. These known types of porous graphite fuel kernels are used to encapsulated fissionable material such as uranium oxide into so called BISO fuel kernels that are commercially available. It is also understood that such commercially available porous graphite kernels have in some cases been coated with one or more layers of graphite and/or pyrolytic carbon, in order to allow gases generated by fission of the nuclear fuel to be absorbed in the surrounding graphite layers and confined therein against release from the fuel elements. Such prior art fuel elements are not suitable for use in high temperature gas reactors that have a sustained operating temperature in excess of 2000.degree. C., because those elements are not capable of retaining molten fissionable material within the fuel kernel until most of the energy available from the fuel has been transferred to the moderating gases of the reactor. Moreover, such prior art fuel elements do not provide a diamond coating or other suitable means for providing a barrier or shield that protects molten fissionable material within the fuel kernels from contact with the high temperature gases used to moderate the nuclear reaction within such ultra high temperature reactors, i.e. those that operate in excess of 2000.degree. C. In nuclear fuel elements made according to the present invention, fissionable material is melted within the pores of graphite fuel elements, thereby to cause the fissionable material to chemically react with the walls of the pores in a manner that localizes and stabilizes the fissionable material within the pores. Such fuel elements are also provided with one or more coatings of pyrolytic carbon or diamond, thereby to form barriers that further localize and stabilize the fissionable material within the pores of the fuel element and also serve as barriers that shield the fissionable material from exposure to the high temperature moderating gases present in high temperature gas reactors. Referring now to FIG. 1 of the drawing, the preferred steps of the method of the invention will be described. In practicing this method one provides a plurality of porous graphite members, such as the commercially manufactured BISO or TRISO nuclear fuel kernels that are readily available. Next, those graphite members are impregnated with a suitable oxidant, such as air or oxygen, and the members are heated to increase their porosity by creating a controlled reaction between the graphite of the members and the oxidant. Next, the graphite members are impregnated with a conventional solution of fissionable fuel material, such as uranyl nitrate and a suitable solvent, such as water. After the members are thus impregnated, the solvent is evaporated to leave the fissionable fuel material deposited within the pores of the members. A next important step of the method of the invention is to heat the graphite members sufficiently to cause the fissionable material to react and then melt with the graphite walls of the pores in the members, thereby to form uranium carbide and to cause the molten fissionable material to be stabilized and localized within the pores of the graphite members. If desired for given applications, at this point in practicing the method of the invention, one may re-impregnated the graphite members one or more times with a solution of fissionable fuel material and solvent, such as that used for the initial impregnation, then the solvent is evaporated from the graphite members. Another optional step at this point in practicing the method is to again heat the members above the melting point of the fissionable material to cause the fissionable material and the graphite walls of the pores to undergo further chemical reaction, thereby to further localize and stabilize the molten fissionable material within the pores. Once a desired level of loading of fissionable materials within the pores is thus achieved, the pores are covered with a layer of graphite to seal the fissionable material therein. Finally, the graphite members are coated with a layer of pyrolytic carbon or diamond so that the coating layers provide barriers that further localize the fissionable material within the pores and also provides a barrier between the fissionable material and the hot moderating gases used in high temperature gas reactors. In certain applications of the invention, it is desirable to deposit one or more additional layers of pyrolytic carbon or diamond over the outer surface of the graphite members or over the earlier-applied coating layers thereon, in order to more fully localize the molten fissionable material within the pores of the graphite member, and to better shield the fissionable material from the moderating gases of an associated reactor. The desired layers of pyrolytic carbon or diamond are deposited by using a suitable conventional vapor deposition process that is controlled to make the layers as suitable barriers against migration of molten fissionable material from the pores of the members. It should be understood that by practicing the method of the invention it is possible to manufacture high temperature nuclear fuel elements in many different configurations. Some of the more preferred configurations for such high temperature fuel elements will now be described. FIG. 2 is a schematic illustration of a basic porous core component for the nuclear fuel element structure of the present invention. As shown in FIG. 2, this basic component comprises a generally spherical member 1 made of a suitable conventional, commercially available porous graphite or carbon. A plurality of irregularly shaped pores in the member 1 are formed to have a desired fuel-carrying volume, by any suitable commercially available process for manufacturing and shaping the pores in such porous graphite or carbon members. In the preferred embodiment of the invention, each of the members 1 is made to be approximately 500 microns in diameter, but it will be understood that the particular outer surface configuration of the member 1 and its size is not critical in practicing the method of the present invention. It should be noted that it is typical in such a member that some of the pores open at the surface of the member, while other pores are closed relative to exposure to the outer surface. Accordingly, the accessibility of the pores to impregnation with fissionable fuel material is often somewhat restricted by the number of open-ended pores within the members. FIG. 3 illustrates this point by showing a graphite member 1 having its pores 2 in which a suitable fissionable fuel material 3 is disposed only in those pores that open to the surface of the member. Several closed-end pores 2A are not accessible to fuel/solvent at the surface of member 2, so those pores 2A remain empty. As noted above with respect to the description of the preferred method steps of the invention, the fissionable fuel material 3 can be stabilized and localized within the pores 2 of member 1 by heating the member 1 sufficiently to melt the fissionable material and to cause it to chemically react with the walls of the pores 2. After one or more such melting steps have been performed in order to localize the fuel within the member 1, the pores are thus made effective to hold the fuel within the pores, through resultant capillary forces and surface tensions forces with the molten fissionable fuel material and the walls of the pores. After a desired number of fuel-loading steps have been performed on the element 1, as discussed in greater detail relative to the method steps described above, a coating of pyrolytic carbon is formed over substantially the entire outer surface of the member 1, as is illustrated in FIG. 4. The coating of pyrolytic carbon 4 forms a kinetic barrier against migration of molten fissionable material from the pores 2; thus, the coating of pyrolytic carbon 4 is effective to further localize and stabilize the fissionable material 3 within the pores 2 of element 1. It is important to note that the coating of pyrolytic carbon 4 is not a porous pyro-carbon structure, which would permit expansion of the graphite member 1, or which would accomodate the gaseous fission products that are generated as the fissionable fuel material is consumed. Instead, the coating 4 is made of a dense, non-porous pyrolytic carbon so that it is effective to prevent the migration of molten fissionable material 3 from the pores 2. In a modified embodiment of the new fuel element 1 of the invention, a coating of diamond 5 is deposited over the entire outer surface of the pyrolytic carbon 4 to act as a further kinetic barrier to the migration of melted fissionable material from the pores 2 and to further act as a barrier against the reaction of reactor moderating gases, such as hydrogen or helium, with the molten fissionable material 3. Another modification of the fuel element of the invention illustrated in FIG. 4 can be achieved by replacing the coating of pyrolytic carbon 4 with a coating of diamond 5 being deposited directly on the graphite member 1, so that the diamond coating 5 would directly seal the pores 2 and also serve as a barrier against reaction between the reactor moderating gases and the molten fissionable material 3 within the member 1. It has been found preferable to form the diamond coating 5 on either the outer surface of the porous graphite or carbon member 1 or on the coating of pyrolytic carbon 4, by use of a conventional controlled vapor deposition process in which hydrogen is present in a concentration greater than about 95% of the gas ambient for the deposited diamond film during the vapor deposition step. That concentration of hydrogen is effective to minimize the presence of graphite in the deposited diamond coating 5. It has also been found that the nuclear fuel element configuration illustrated in FIG. 4 can be further modified by depositing a layer of porous graphite carbon over the entire outer surface of the coating of pyrolytic carbon 4, between it and the diamond coating 5. A portion (only) of such a layer of a porous graphite carbon 6 is illustrated in FIG. 4. By using such alternate layers of different forms of carbon, the successive layers are made more effective to form a series of kinetic barriers to further localize and stabilize molten fissionable material 3 within the pores 2. In the preferred embodiment of the nuclear fuel element structure illustrated in FIG. 4, the fissionable material 3 comprises a composition of uranium or plutonium carbide or nitrate and the porous graphite member 1 is either selected or suitably modified by the oxidation steps of the method of the invention to assure that at least some of the pores do extend to the outer, generally spherical surface of the member 1 and are greater in length than a radius of member 1, as is clearly illustrated in FIG. 4. The thicknesses of the respective coatings 4 and 5 illustrated in FIG. 4 must be effective to form barriers against migration of the fissionable fuel material from the pores 2, and the outer diamond coating 6 must be effective to shield the fissionable material from the reactor moderating gases; thus, each of the coatings 4 and 5 should be made at least 25 microns thick, up to about 25 mils thick, measured in a radial direction. The diamond coating 6 is preferably made 25 microns to 5 mils thick. Another preferred configuration of nuclear fuel element made according to the invention is one in which the fuel elements are made as relatively thin filaments or fibers that are capable of performing at temperatures up to the sublimation temperatures of graphite, i.e. at temperatures greater than 3300.degree. C. Suitable carbon filaments or fibers for practicing this form of the invention are commercially available in both porous and solid graphite filament form. There is illustrated in FIG. 5, in transverse cross section along a diameter of such a porous carbon or graphite filament member 1A a generally circular (or cylindrical) configuration for the filament, but it will be recognized that such filaments may have other cross section configurations without departing from the scope of the present invention. A plurality of pores 2A that extend into the filament 1A from its outer surface are filled with fissionable material 3A in a manner similar to that used for filling the pores 2 of the generally spherically carbon or graphite fuel elements illustrated in FIGS. 3-4. In addition to the fissionable material that is localized within the pores 2A, a layer of fissionable material 3B is deposited around the graphite or carbon filament element 1A. In practicing the method of the invention to make such a filament type fuel element it has been found that a carbon filament can be suitably heated by passing electric current through it, after it is suitably electrically connected between conventional commercially available terminals, which in turn are operatively connected to a conventional source of electric power. The graphite filaments can thus be either partially or totally converted to a nuclear fuel element by controlling the current, time and pressure of an ambient gas or vapor environment of a suitable fissionable material such as uranium hexalflouride, which is made to surround the filament during the fuel impregnation step. By thus suitably controlling the heat transfer geometry within such a conventional furnace, uranium or other suitable fissionable material is deposited within the pores 2A and if desired the deposition is continued to build up a surrounding coating 3B of fissionable material, as illustrated in FIG. 5. By controlling current passed through the filament thereby to regulate the temperature of the graphite filament 1A, the fissionable material 3A within the pores 2A is melted sufficiently to cause it to react with the graphite walls definining the pores 2A, thereby to localize and stabilize the molten material within the pores. After that reaction, the gas used to deposit the fissionable fuel material in and around the graphite film element 1A is removed as an ambient for the filament and appropriate alternative gases are used to deposit either pyrolytic carbon or diamond, such as the coating layer 4A illustrated in FIG. 5. As was the case with the type of nuclear fuel element illustrated in FIGS. 3-4, a multiple-layer fuel element can be made in a filament configuration, as is shown in FIG. 6. In FIG. 6, a hollow porous carbon or graphite member 1A' having pores 2A' therein is impregnated with a suitable fissionable material 3A' and is surrounded by additional fissionable material 3B' that is coated on both the interior and exterior surfaces of the member 1A'. By suitably raising and lowering the temperature of the filament 1A', for example by selectively regulating the electric current that is passed through it, and by changing the gas composition and the pressure of the furnace ambient surrounding the filament during such heating, predetermined and controlled layers of graphite, pyrolytic carbon, and diamond can be formed on the filament, generally in the manner noted above. Thus, as is shown in FIG. 6, the fissionable fuel coating 3B' on the filament 1A' is coated with a layer of pyrolytic carbon 4A', which in turn is surrounded by another layer of fissionable fuel material 3C', which in turn is coated with another layer of pyrolytic carbon 4B' which acts as a kinetic barrier against migration of the fissionable material 3C' through the pyrolytic carbon coating 4B' according to the present invention. It should understood that although only a portion of a diamond coating 5A' is illustrated, this coating is formed to completely surround the outer surface of the pyrolytic carbon layer 4B', thereby to form a kinetic barrier against migration of fissionable fuel material from the fuel element, and also to form a barrier between the moderating hydrogen or helium gas used in a high temperature gas reactor, and the fissionable fuel material within the fuel element. One advantage of the filament type fuel element illustrated in FIGS. 5 and 6 is that they are sufficiently flexible to enable the individual filaments to be twisted to form a thicker bundle of such elements, which in turn can be deposited in a graphite housing. For example, as is shown in FIG. 7, a plurality of such fissionable fuel elements 10 are schematically shown surrounded by a body of graphite 11, which preferably is pyrolytic carbon that forms yet another barrier against migration of the fissionable material from the elements 10. Of course, the cross section configuration of individual fuel element 10 can be made of any desired multiple layer configuration, such as the configurations shown in FIGS. 5 and 6. A desirable feature of such a multi-filament fuel element is that it permits a number of different fissionable fuel materials to be used in selected combinations within a single multi-fuel element such as the graphite cylinder 11 illustrated in FIG. 7. It should be understood that, according to the present invention, the fissionable fuel materials used in making a multi-fuel element bundle, such as that shown in FIG. 7, can be stabilized and localized within the fuel member 11 by heating it to melt the fissionable material and cause is to react with the graphite that surrounds the fissionable material, thereby to help prevent the fissionable material from migrating out of the fuel element. As noted above, a diamond coating could be provided over the exterior surface of the element 11 to form a further barrier against migration of fissionable fuel material from the combined fuel element 11, as well as to prevent the fissionable material from reacting with the reactor gases. To illustrate such a modification a simpler form of flexible fuel filament is shown in FIG. 8. In this modification a hollow porous graphite filament 12 has a fissionable fuel material 13 impregnated within its pores (not shown) and built up on its inner cylindrical surface. A diamond coating 14 is formed by a conventional vapor deposition process over the outer surface of the filament 12. Referring, again to FIG. 7, it should also be noted that when a plurality of such filaments 10 are positioned adjacent to one another to form a true element bundle, i.e. without added graphite of the member 11 between the filaments (10) as shown in FIG. 7, the juxtaposed surfaces of the respective filaments 10 provide additional surface barriers that further serve to localize and stabilize molten fissionable fuel materials within the respective filaments 10. It will be appreciated that a bundle of filament, fuel-containing elements 10, such as those shown in FIG. 7, may be formed by either twisting individual filaments 10 together, or by pressing a body of graphite material, such as the material 11, around individual filaments to form a larger cylindrical multi-filament fuel element of the type shown in FIG. 7. Alternatively, a body of graphite (11) could be bored to form passageways for accepting the fuel filaments (10). From the foregoing description of the invention it will be apparent to those skilled in the art that various further modifications and alternative embodiments of it may be developed without departing from the scope of the invention; thus, it is my intention to encompass within the following claims the true limits of the invention.
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	claims	1. A method for process monitoring, the method comprising the steps of:receiving a sample that defines a cavity, said sample made of at least a first material;determining at least one characteristic of the cavity;receiving a sample that comprises a processed cavity filled with a second material;directing a beam of charged particles towards the sample, so as to induce X-ray emission from a first portion of the sample, said first portion at least partially overlaps the processed cavity;detecting X-ray emitted from said first portion; andproviding an indication about the process in response to detected X-ray emission from the first portion and the at least one determined characteristic of the cavity. 2. The method of claim 1 wherein the step of determining comprises measuring at least one dimension of the cavity at one or more locations. 3. The method of claim 1 wherein the step of determining comprises estimating a volume of the cavity. 4. The method of claim 1 wherein the step of determining comprises scanning a portion of the cavity with a radiation beam. 5. The method of claim 1 wherein the cavity is processed by polishing a previously filled processed cavity. 6. The method of claim 1 wherein the second material is optically oblique. 7. The method of claim 1 wherein the indication reflects the presence of voids within the first portion. 8. The method of claim 1 wherein the indication reflects a shape of the filled cavity. 9. The method of claim 1 wherein the indication reflects the thickness of the processed cavity at various locations. 10. The method of claim 1 wherein the sample defines multiple cavities that are processed to provide multiple processed cavities, each cavity associated with a different portion of the sample. 11. The method of claim 10 wherein the steps of directing a beam and detecting emitted X-rays are repeated for each of the multiple processed cavities. 12. The method of claim 11 wherein the indication about the process is responsive to the detected X-ray emission from the portions associated with each processed cavity and a determined characteristic of at least one cavity. 13. The method of claim 10 further comprising providing a map of the sample indicating X-ray emission measured in response to the detected X-ray emission from the multiple portions of the sample. 14. The method of claim 10 further comprising a step of locating the multiple processed cavities. 15. The method of claim 14 wherein locating a processed cavity comprises acquiring an image of an estimated vicinity of the processed cavity and processing the image to locate the processed cavity. 16. The method of claim 15 wherein the image is acquired by scanning the sample within an acquisition window. 17. The method of claim 16 wherein a processed cavity is scanned within a scanning window that is smaller than the acquisition window. 18. The method of claim 1 wherein the indication is further responsive to a reference parameter. 19. The method of claim 18 wherein the reference parameter is responsive to at least one measurement of at least one other processed cavity. 20. The method of claim 18 wherein the reference parameter is responsive to an estimated X-ray emission. 21. The method of claim 1 further comprising a step of changing a characteristic of the beam of charged particles to provide a changed beam. 22. The method of claim 21 further comprising directing the changed beam towards the sample, so as to induce X-ray emission from a second portion of the sample, said second portion at least partially overlaps the processed cavity; and detecting X-ray emitted from said second portion. 23. The method of claim 22 wherein the indication about the process is further responsive to detected X-ray emission from the second portion. 24. A method for process monitoring, the method comprising the steps of:receiving a sample comprising at least two materials;scanning an area of the sample such as to induce X-ray emission from a first portion of the sample;detecting X-ray emitted from said first portion; andproviding an indication about the process in response to the detected x-ray emission, wherein the step of providing comprises applying a quantitative analysis correction technique on the detected X-ray emission. 25. The method of claim 24 wherein the technique is ZAF analysis. 26. The method of claim 24 wherein the estimated detected radiation is responsive to detected X-ray measurements from at least one other area of the sample. 27. The method of claim 24 wherein the estimation is responsive to an acceleration voltage applied on a beam that scans the area. 28. The method of claim 24 wherein the indication reflects a presence of a void within the first portion. 29. The method of claim 24 wherein the indication reflects a shape of the area. 30. The method of claim 24 further comprising a step of providing an estimate of a characteristic of a reference object and wherein the indication is responsive to said estimated characteristic. 31. The method of claim 30 wherein the characteristic is a thickness of the reference object. 32. The method of claim 30 wherein the characteristic is measured at a location that is selected such as to provide an indication of the characteristic. 33. The method of claim 30 wherein the location is selected in response to a process characteristic. 34. The method of claim 30 wherein the location is positioned substantially at a center the reference object. 35. The method of claim 30 wherein the reference object is large. 36. A system for process monitoring, the system comprising:means for determining at least one characteristic of a cavity defined by a sample made of at least a first material;means for directing a beam of charged particles towards the sample so as to induce X-ray emission from a first portion of the sample, said first portion at least partially overlaps a processed cavity; whereas the cavity was processed to provide the processed cavity filled with a second material;at least one detector for detecting X-ray emitted from said first portion; anda processor, coupled to the at least one detector, for providing an indication about the process in response to detected X-ray emission from the first portion and the at least one determined characteristic of the cavity. 37. The system of claim 36 wherein the indication reflects the presence of voids within the first portion. 38. The system of claim 36 wherein the indication reflects a shape of the filled cavity. 39. The system of claim 36 wherein the indication reflects the thickness of the processed cavity at various locations. 40. The system of claim 36 wherein the sample defines multiple cavities that are processed to provide multiple processed cavities, whereas each processed cavity is associated with a different portion of the sample. 41. The system of claim 40 wherein the means for directing a beam are further adapted to direct the beam toward each of the multiple processed cavities and to measure the emitted X-ray from each processed cavity. 42. The system of claim 41 wherein the processor is adapted to provide an indication about the process in response to the detected X-ray emission from the portions associated with each processed cavity and a determined characteristic of at least one cavity. 43. The system of claim 41 wherein the processor is further adapted to provide a map of the sample indicating X-ray emission measured in response to the detected X-ray emission from the multiple portions of the sample. 44. The system of claim 41 wherein the system is further adapted to locate the multiple processed cavities. 45. The system of claim 36 wherein the indication is further responsive to a reference parameter. 46. The system of claim 36 further capable of changing a characteristic of the beam of charged particles to provide a changed beam. 47. The system of claim 46 further adapted to direct the changed beam towards the sample, so as to induce X-ray emission from a second portion of the sample, said second portion at least partially overlaps the processed cavity; and to detect X-ray emitted from said second portion. 48. The system of claim 47 wherein the indication about the process is further responsive to detected X-ray emission from the second portion. 49. A system for process monitoring, the system comprising:means for scanning an area of the sample such as to induce X-ray emission from a first portion of the sample;at least one detector for detecting X-ray emitted from said first portion; anda processor for applying a quantitative analysis correction technique on detected X-ray emission and in response provide an indication about the process. 50. The system of claim 49 wherein the indication reflects a presence of a void within the first portion. 51. The system of claim 49 wherein the indication reflects a shape of the area. 52. The system of claim 49 further adapted to provide an estimate of a characteristic of a reference object and wherein the indication is responsive to said estimated characteristic.
	description	This application is a divisional of application Ser. No. 10/742,778 filed. Dec. 23, 2003 and claims the benefit of priority from the prior Japanese Patent Application No. 2002-382394, filed Dec. 27, 2002, the entire contents of which are incorporated herein by reference. 1. Field of the Invention The invention relates to lithography using a charged particle beam. More specifically, the invention relates to a stage phase measurement method for a charged particle beam exposure apparatus for measuring the phase of a mask stage coordinate system for a specimen stage coordinate system of a charged particle beam exposure apparatus; a demagnification measurement method for a charged particle beam exposure apparatus for measuring the demagnification for image projection onto a mask specimen surface; a control method for a charged particle beam exposure apparatus for performing control corresponding the measured phase and demagnification; and a charged particle beam exposure apparatus. 2. Description of the Related Art With increasingly fined semiconductor devices, studies and research are being made regarding charged particle beam exposure apparatuses for exposure patterns. Demagnification lenses and objective lenses are used to demagnify and transferring a mask pattern onto a specimen. The mask pattern is demagnified by these lenses and the pattern is rotated by a magnetic field, so that the phase of the pattern to be transferred onto the surface of the specimen is varied concurrently with deflection in demagnification. The apparatus is designed by taking both the rotation and the demagnification into account, and the apparatus is designed so that the rotation is performed at a desired demagnification. Practically, however, design errors and manufacture errors disable obtaining the condition concurrently allowing the desired demagnification and the desired rotation to be exhibited. A process for measuring the demagnification and pattern rotation angle is disclosed in Jpn. Pat. Appln. KOKAI Publication No. 7-22349. A rotational error of the pattern is correctable by using a rotation stage that carries the mask. However, since no means is provided to correct the movement direction of a mask stage X and the movement direction of a mask stage Y, the system phase of the mask stage coordinate system remains mismatched with the specimen stage coordinate system. Because of assembly errors, design errors, and lens system adjustment errors, the mask stage coordinate system has errors for the specimen stage coordinate system; and generally, it does not have means for adjusting the errors. While an XY mask stage should be mounted to one more θ stage to adjust the phase of an XY mask stage, since the construction is thereby complexed and free space in an electrooptical housing is insufficient, it is difficult to mount the XY mask stage. When moving a desired mask pattern with the mask stage to the vicinity of the beam, if such errors as those described above are zero, the movement position can be determined in accordance with pattern design values. However, a problem arises in that an accurate movement position of the pattern cannot be known, so that accurate movement cannot be implemented. Regarding a demagnification measurement method, using a design distance D between two opening portions provided in the mask and a distance d between individual beam specimen surface positions formed in the opening portions, the demagnification has been obtained by way of “demagnification M=d/D”. However, errors such as those occurring in the manufacture of the opening portions and distortion undesirably influence the calculation result. In a case where the manufacture error is 50 nm and the distance between the opening portions is 500 μm, the case results in causing an error of 0.01% (50 nm/500 μm×100). When performing scan-exposure of a 300 μm mask pattern by using the demagnification, there arises the problem of causing an image-dimensional error of as large as 30 nm (i.e., 300 μm×0.01%=30 nm). Further, a problem arises in that an accurate pattern cannot be imaged onto the specimen since no method is available to measure the phase of the mask stage coordinate system with respect to the problem of demagnification measurement errors and the specimen stage coordinate system. A demagnification measurement method for a charged particle beam exposure apparatus, according to an aspect of the present invention, comprises: measuring a first stage position of a mask stage of the charged particle beam exposure in accordance with a mask stage coordinate system with an opening portion of a mask placed on the mask stage being situated in a first opening position; irradiating a first charged particle beam to a first irradiation position on a surface of a specimen through the opening portion of the mask, the first charged particle beam being shaped through the opening portion and then passing through an objective lens system; measuring the first irradiation position in accordance with a specimen stage coordinate system; moving the mask stage to a second stage position to situate the opening portion of the mask in a second opening position different from the first opening position; measuring the second stage position of the mask stage in accordance with the mask stage coordinate system; irradiating a second charged particle beam to a second irradiation position on the surface of the specimen through the opening portion of the mask moved together with the mask stage, the second charged particle beam being shaped through the opening portion situated in the second opening position and then passing through an objective lens system; measuring the second irradiation position in accordance with the specimen stage coordinate system; and calculating a demagnification of the objective lens system from the first and second stage positions and the first and second irradiation positions. A stage phase measurement method for a charged particle beam exposure apparatus, according to another aspect of the present invention, comprises: measuring a rotation angle of a pattern of a charged particle beam shaped through a mask placed on a mask stage of the charged particle beam exposure apparatus and then irradiated on a surface of a specimen through an objective optical system; correcting rotation of the pattern by rotating the mask corresponding to the measured rotation angle; measuring a first stage position of the mask stage in accordance with a mask stage coordinate system after correcting the rotation with an opening portion of the mask being situated in a first opening position; irradiating a first charged particle beam to a first irradiation position on the surface of the specimen through the opening portion of the mask, the first charged particle beam being shaped through the opening portion and then passing through an objective lens system; measuring the first irradiation position in accordance with a specimen stage coordinate system; moving the mask stage to a second stage position to situate the opening portion in a second opening position different from the first opening position; measuring the second stage position of the mask stage in accordance with the mask stage coordinate system; irradiating a second charged particle beam to a second irradiation position on the surface of the specimen through the opening of the mask moved together with the mask stage, the second charged particle beam shaped through the opening portion situated in the second opening position and then passing through the objective lens system; measuring the second irradiation position in accordance with a specimen stage coordinate system; and calculating a phase difference between the specimen stage coordinate system and the mask stage coordinate system from the first and second stage positions and the first and second irradiation positions. A control method for a charged particle beam exposure apparatus, according to another aspect of the present invention, comprises: measuring a first stage position of a mask stage of the charged particle beam exposure apparatus in accordance with a mask stage coordinate system with an opening portion of a mask placed on the mask stage being situated in a first opening position; irradiating a first charged particle beam to a first irradiation position on a surface of a specimen through the opening portion of the mask, the first charged particle beam being shaped through the opening portion situated in the first opening position and then passing through an objective lens system of the exposure apparatus; measuring a first irradiation position in accordance with a specimen stage coordinate system; moving the mask stage to a second stage position to situate the opening portion in a second opening position different from the first opening position; measuring the second stage position of the mask stage in accordance with the mask stage coordinate system; irradiating a second charged particle beam to a second irradiation position on the surface of the specimen through the opening portion of the mask moved together with the mask stage, the second charged particle beam being shaped through the opening portion situated in the second opening position and then passing through the objective lens system; measuring the second irradiation position in accordance with a specimen stage coordinate system; and obtaining a demagnification of the objective lens system from the first and second stage positions and the first and second irradiation positions; adjusting the demagnification of the objective lens system corresponding to the obtained demagnification; measuring a rotation angle of a pattern of the charged particle beam shaped through the mask and then irradiated on the surface of the specimen via the objective optical system, after the adjusting; correcting the rotation of the pattern by rotating the mask corresponding to the measured rotation angle; measuring a third stage position of the mask stage in accordance with a mask stage coordinate system after correcting the rotation with the opening portion of the mask being situated in a third opening position; irradiating a third charged particle beam to a third irradiation position on the surface of the specimen through the opening portion situated in the third opening position, the third charged particle beam being shaped through the opening portion situated in the third opening position and then passing through an objective lens system; measuring the third irradiation position in accordance with a specimen stage coordinate system; moving the mask stage to a fourth stage position to situate the opening portion in a fourth opening position different from the third opening position; measuring the fourth stage position of the mask stage in accordance with the mask stage coordinate system; irradiating a fourth charged particle beam to a fourth irradiation position on the surface of the specimen through the opening portion situated in the fourth opening position, the fourth charged particle beam being shaped through the opening portion situated in the fourth opening position and then passing through the objective lens system; measuring the fourth irradiation position in accordance with a specimen stage coordinate system; and obtaining a phase difference between the specimen stage coordinate system and the mask stage coordinate system from the third and fourth stage positions and the third and fourth irradiation positions; and moving the mask stage by correction in accordance with the phase difference. A charged particle beam exposure apparatus according to another aspect of the present invention, comprises: a radiating unit configure to radiate a charged particle beam; an XY mask stage on which a mask having an opening is placed and which moves the mask stage in X and Y directions of a mask stage coordinate system; a mask stage measuring unit configured to measure a position of the XY mask stage in accordance with the mask stage coordinate system; a deflector which deflects the charged particle beam and changes the position of the charged particle beam on a surface of the mask; an objective lens system which demagnifies a pattern of the charged particle beam shaped through the mask and irradiates the specimen with the charged particle beam; a specimen stage on which the specimen is placed and which moves the specimen in X and Y directions of a specimen stage coordinate system; an objective deflector which deflects the charged particle beam and changes the position of the charged particle beam on a surface of the specimen; an irradiation position measuring unit configure to measure an irradiation position of the charged particle beam on the surface of the specimen in accordance with the specimen stage coordinate system; and a demagnification measuring unit configure to measure a demagnification of the objective lens system on the basis of two positions of the XY mask stage measured at different opening positions respectively and a position of the charged particle beam on the surface of the specimen that was shaped through the opening of each of the opening positions. A charged particle beam exposure apparatus according to another aspect of the present invention, comprises: a radiate unit configure to radiate a charged particle beam; an XY mask stage on which a mask having an opening is placed and which moves the mask in X and Y directions of a mask stage coordinate system; a θ mask stage which rotates the mask in an XY plane of the mask stage coordinate system; an opening position measuring unit configure to measure a position of the opening in accordance with the mask stage coordinate system; a deflector which deflects the charged particle beam and changes the position of the charged particle beam on a surface of the mask; an objective lens system which demagnifies a pattern of the charged particle beam shaped through the mask and irradiates a specimen with the charged particle beam; a specimen stage on which the specimen is placed and which moves the specimen in X and Y directions of a specimen stage coordinate system; an objective deflector which deflects the charged particle beam and changes the position of the charged particle beam on a surface of the specimen; an irradiation position measuring unit configure to measure an irradiation position of the charged particle beam on the surface of the specimen in accordance with the specimen stage coordinate system; a rotation angle measuring unit configure to measure a rotation angle of the pattern of the charged particle beam in the objective lens system; a phase measuring portion configure to measure a phase of the mask stage coordinate system with respect to the specimen stage coordinate system based on a position of the XY mask stage measured at two opening positions respectively and a position of the charged particle beam on the surface of the specimen that was shaped through the opening of each of the opening positions; and a driving unit configure to drive the XY mask stage and the θ mask stage corresponding to the measured phase. An embodiment according to the present invention will be described herein below with reference to the drawings. FIG. 1 is a schematic configuration diagram showing an electron beam exposure apparatus according to the embodiment of the present invention. A beam emitted from an electron gun 1 is imaged through an illumination lens 2, a projection lens 19, and a demagnification lens (objective lens system) 8, and is finally imaged on a main surface of an objective lens (objective lens system) 9. An image of a first shaping aperture 4 is formed onto a mask 6, and an image thus formed is created on the specimen surface through the demagnification lens 8 and the objective lens 9. An opening portion 21 is provided in the first shaping aperture 4. The dimensional shape of the opening portion 21 is a rectangle having one side of 80 μm, for example. The electron beam emitted from the electron gun 1 can be deflected through a blanking deflector 3, and the beam position on the first shaping aperture 4 can thereby be changed. Referring to FIG. 2, the mask 6 is mounted over a θ stage 20, and the θ stage 20 is mounted on an X stage 14 and a Y stage 15, whereby the mask 6 can be moved. The mask 6 is moved by the X stage 14 and the Y stage 15 in X and Y directions. The positions of the X stage 14 and the Y stage 15 are under positional control of a laser measurement apparatus (laser interferometer) 31. An opening portion 22 is provided in the mask 6, as shown in FIG. 2. In accordance with programs stored in a storage medium 35, a CPU 34 acquires the position of the opening portion 22 from the measurement result of the laser measurement device 31. The dimensional shape of the opening portion 22 is smaller than the size of the first shaping aperture image formed on the mask 6. The dimensional shape of the opening portion 22 is a rectangle having one side of 40 μm, for example. The electron beam shaped through the opening portion 21 of the first shaping aperture can be deflected through a shaping deflector 5, and the beam position on the mask 6 can thereby be changed. The beam passed through the objective lens 9 can be deflected by an objective deflector 18. A marking table 10 is provided on an XY specimen stage 11 and is movable in the X and Y directions in the specimen stage coordinate system. The position of the XY specimen stage 11 is under positional control of a laser measurement device (laser interferometer) 32. As shown in FIG. 3, a cross mark 17 provided on the marking table 10 is made from a beam-reflecting material different from a material of a base 24. For example, the base 24 is made of silicon, whereas the mark 17 is made of a material such as gold or tungsten, for example. The beam position on the marking table 10 can be changed by the objective deflector 18. A beam detector 23 detects electrons reflected from the marking table 10 and secondary electrons. A function is provided that moves the mark 17 to the optical axis position, scans the electron beam to be projected onto the mark 17 by using the objective deflector 18, and then detects the irradiation position of the electron beam in accordance with the distance between the mark and the electron beam, which has been obtained through calculation performed by taking a signal detected by the beam detector 23 into a mark signal processor 33 and a stage position measurement value of the laser measurement device 32. The CPU 34 executes the above-described function in accordance with programs stored in the storage medium 35. In FIG. 1, reference numeral 17 denotes an objective aperture, reference numeral 12 denotes a lens imaging system, and reference numeral 13 denotes a shaped-image imaging system 13. A mask stage phase measurement method and an objective-lens-system demagnification measurement method according to the present embodiment will now be described hereinbelow by using FIGS. 4A, 4B, and 5. The mask stage phase measurement and the objective-lens-system demagnification measurement are executed by the CPU 34 in accordance with programs stored in the storage medium 35. Also, control of the X, Y, and θ mask stages 14, 15, and 20 corresponding to the measurement results is executed by the CPU 34 in accordance with programs stored in the storage medium 35. When measuring a mask stage phase, a rotation angle θmp of the mask pattern formed through the demagnification lens 8 and the objective lens 9 is preliminarily measured. The θ stage 20 is driven in accordance with the measurement result to bring the mask pattern into to the state in which it is not rotated on the specimen. A method of measuring a mask-pattern rotation angle as θmp is described in, for example, Jpn. Pat. Appln. KOKAI Publication No. 7-22349. However, the measurement of the rotation angle θmp is not necessary in the event of obtaining only the demagnification. The opening portion 22 on the mask 6 is moved to a position A by using the shaping deflector. The positions of the X mask stage 14 and the Y mask stage 15 are measured by a laser measurement device in accordance with the mask stage coordinate system, and the position A of the opening portion 22 is measured from the results thereof. The electron beam shaped through the first shaping aperture opening portion 21 is deflected by the shaping deflector 5 to the opening portion 22 on the mask. The electron beam is deflected to a position where the opening portion 22 is covered overall, as shown in FIG. 4A. The electron beam shaped through the opening portion 22 arrives at a position “a”, as shown in FIG. 4B. The position “a” is measured by a mark scan process performed in accordance with the specimen stage coordinate system. The mark scan process is described in, for example, reference document (S. Nishimura: Jpn. J. Appl. Phys. Vol. 36 (1997), pp. 7517-7522: Evaluation of Shaping Gain Adjustment Accuracy Using Atomic Force Microscope in Variably Shaped Electron-Beam Writing Systems) and Jpn. Pat. Appln. KOKAI Publication No. 10-270337. Subsequently, the opening portion 22 of the mask 6 is moved to a position B (FIG. 4A). The positions of the X mask stage 14 and the Y mask stage 15 are measured by a laser measurement device in accordance with the mask stage coordinate system, and the position B is measured from the results thereof. The electron beam is deflected by the shaping deflector to the opening portion 22 in the position B. The beam shaped through the opening portion 22 is then irradiated to a position b (FIG. 4B) on the specimen. In a manner similar to the above, the position b is measured by the mark scan process in accordance with the specimen stage coordinate system. The electron beam is irradiated to the two positions without altering the settings of the demagnification lens 8, the objective lens 9, and the objective deflector 18. The distance between the position A and the position B of the opening portion on the mask 6 is represented by L. Likewise, the distance between the beam position “a” and the beam position b of each specimen surface is represented by l. In this case, the relationship can be expressed as “demagnification η=1/L.” In addition, the phase difference between a line segment connecting between the position A and the position B and an Xm axis of the mask stage coordinate system is represented by θ1. Likewise, the phase difference between a line segment connecting between the position “a” and the position b and the X axis of the specimen stage coordinate system is represented by θ2. In this case, a phase θ of the mask stage coordinate system for the specimen stage coordinate system can be expressed as “θ2 −θ1 ” (FIG. 5). The phase differences θ1 and θ2 are thus obtained based on the Xm axis and the X axis. However, the phase differences θ1 and θ2 may be obtained based on a Ym axis and the Y axis. Still alternatively, the phase differences θ1 and θ2 may be obtained based on straight lines having the same tilts in the individual coordinate systems. In addition, according to the above description, the opening positions A and B are individually obtained. However, only the positions of the X mask stage 14 and the Y mask stage 15 may be measured by the laser measurement device in the state in which the opening portions are individually situated in the opening positions A and B. The positional relationship between the two opening positions can be known and the distances and phase differences can be obtained from the X mask stage 14 and the Y mask stage 15 in the individual opening positions. The positions of the X mask stage 14 and the Y mask stage 15 can be accurately obtained. In the present embodiment, the measurement is performed by way of measurement of the positions of the X mask stage 14 and the Y mask stage 15, so that the measurements are each obtained with a measurement accuracy of 1 nm or less (accuracy of an actual measurement device recently used). Accordingly, also the measurement accuracy of the distance L of each of the positions A and B is 1 nm or less. When the distance L is 500 μm, the error is 50 nm in the conventional case. However, in the present invention, the measurement can be implemented with the accuracy of 1 nm or less, so that the demagnification measurement error is 0.0002% (1 nm/500 μm×100). Therefore, in the case where a pattern of 300 μm is imaged on the mask, a linewidth accuracy or positional accuracy of 0.6 nm (i.e., 300 μm×0.0002×0.01=0.6 nm) can be implemented. Where the measurement accuracy of the irradiation position “a” and the irradiation position b is 1 nm (measurement accuracy of a recent exposure apparatus) and the distance is 50 μm, the phase measurement error is 1/50,000 rad (=0.02 mrad). Where the movement amount of the mask stage is 100 mm, the difference at both ends is as extremely small as 2 μm (100 mm×0.02/1,000). As such, the positional movement accuracy of the pattern on the mask 6 is exhibited with a high value of 2 μm. Further, since the differing phase is corrected and the mask stage is moved, when exposure is performed while the mask stage is being moved, the overall deflection range of Y of the shaping deflection becomes effectively usable as a scan width. A control method for the charged particle beam exposure apparatus, which is configured by combining the above-described demagnification measurement and the stage phase measurement will now be described hereinbelow with reference to FIG. 6. Using the method described above, processing is performed to measure a demagnification η (step S101). Then, the measured demagnification η is compared with a desired demagnification η0 (step S102). If the result is not η=η0, the lens system is adjusted so that the desired demagnification can be obtained (step S103). If the measured demagnification η has become the demagnification η0, processing proceeds to next step S104. The arrangement may be such that if a required demagnification has reached an allowable error range, processing shifts to next step S104. Using a well known process, processing is performed to measure a rotation angle θmp of a pattern of an electron beam that has been shaped through a mask and has traveled through the objective lens system (step S104). The process to be used to measure the rotation angle θmp is selected from those of the type that does not rely on the phase difference between the mask stage coordinate system and the specimen stage coordinate system. Then, the θ stage 20 is driven corresponding to the rotation angle θmp, and the rotation of the pattern is thereby corrected (step S105). Subsequently, using the above-described method, processing is performed to measure a phase θ of the specimen stage coordinate system for the mask stage coordinate system (step S106). When moving the mask stage, after correction is made corresponding to the phase θ, and the mask stage is moved (step S107). Where the mask pattern coordinate system is based on (Xm, Ym) and the mask stage coordinates are based on (X, Y), moving the mask stage to satisfy the following relationship enables the mask stage to be moved in conformity with the phase of the specimen stage:ΔX=ΔXm×cos θ+ΔYm×sin θΔY=ΔYm×cos θ−ΔXm×sin θ By performing the movement correction of the XY mask stage, the mask can be moved to an accurate position. Then, by performing adjustment of the demagnification, correction of the rotation angle of the pattern, and movement correction of the mask stage according to the phase θ, the pattern can be accurately imaged on the specimen. The present invention is not limited to the embodiment described above. While having been described by reference to the exemplified electron beam exposure apparatus, the present invention can be adapted also to an ion beam exposure apparatus. In addition, while the stage position is measured by the laser interferometer, there is no limitation thereto; and any other devices may be used as long as they are capable of defining the stage coordinates with high accuracy. The process of measuring the irradiation position of the electron beam is not limited to the mark scan process. For example, as shown in FIG. 7, a process is available in which scan is performed with an electron beam over a mark M sized smaller than a scan range, and the center of gravity of a screen-image object is obtained to thereby measure the beam position. The mark M may be arbitrary, as shown in FIG. 7. A range R larger than the mark M is beam-scanned to thereby obtain data of the screen-image object. The beam position can be obtained by obtaining the data of the screen-image object. The present invention can be practiced by making various other changes without departing from the scope of the invention. Additional advantages and modifications will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents.
	description	This application claims the benefit of U.S. Application No. 60/700,856, filed Jul. 19, 2005, which is incorporated herein by reference in its entirety. The present invention relates to reactor vessel auxiliary equipment and, more particularly, to an assembly incorporating multiple systems disposed over a pressurized water reactors. In a typical commercial nuclear power plant such as a pressurized water reactor (“PWR”), a number of components and systems are installed on or directly over the reactor vessel closure head. These components and systems may include one or more of the following: a control element drive mechanism (“CEDM,” also referred to as a control rod drive mechanism); a cooling system; a lift rig for the reactor vessel closure head; CEDM seismic restraints; and a CEDM missile shield. The components and systems are typically designed and installed individually to perform designated functions during plant operation. It is well known that removal and subsequent re-installation of the reactor vessel closure head, including the requisite removal of various components disposed about the reactor vessel, is an expensive and time-consuming process. During refueling of the reactor, the installed components are generally disassembled from the reactor, removed and stored, to provide access to the reactor vessel closure head, so that the vessel head may be removed from the reactor vessel. The disassembled components are placed in designated storage areas, generally inside the reactor containment. Typically, in a commercial nuclear power plant, a lengthy series of steps or detailed procedures must be followed to safely remove external such equipment before the reactor vessel closure head is removed from the reactor vessel. The procedures that are performed prior to detensioning the reactor vessel closure head studs will generally include some or all of the following: Removal and storage of heavy concrete missile shields; Removal and storage of the CEDM cooling ducts; Removal of the seismic restraints; Disconnecting and storage of the CEDM power and rod position indicator cables; Installation of the reactor head lifting rig tripod; Removal of the cable trays and cables that extend from the reactor vessel closure head to the operating deck or walls; Disconnecting the heated junction thermocouples, nuclear steam supply system instrumentation, monitoring system cables, and reactor head vent lines; and Installation of temporary radiation shield blankets around the vessel closure head area. The procedure also requires that the nuts and washers be removed from the reactor vessel closure head and placed in storage racks during preparation for refueling. The storage racks are then removed from the refueling cavity and stored at convenient locations inside containment prior to reactor vessel closure head removal and refueling cavity flooding. After refueling and any other desired servicing, the reactor vessel closure head is replaced, and the components and systems are reassembled, generally by reversing the steps mentioned above. Each procedure in the refueling process contributes significantly to the total cost associated with refueling the reactor. The total costs include costs associated with personnel time required to perform the refueling, power plant down time and consequent loss of electricity production, radiation exposure to personnel, and risks and costs associated with potential human errors. In addition, the various components that must be removed for refueling activities require a large amount of the limited storage space available inside containment and raise the risk of inadvertent contamination of work and storage areas. Concepts and designs for integrating some of the reactor vessel closure head systems into a modular integrated head design have been proposed. For example, in U.S. Pat. No. 4,678,623 to Malandra et al., a head assembly is disclosed wherein vertical lift rods are attached to the reactor vessel lifting lugs and a missile shield, seismic support platform, CRDM cooling system, and lift rig are supported by the lift rods above the reactor vessel closure head. Because most or all of the modular head assembly taught by Malandra et al. is supported by the lift rods, however, very large loads are concentrated at the clevis connection at the reactor vessel closure head lifting lugs, which may cause damage to the lifting lugs and/or the body of the reactor vessel closure head. In addition, very heavy components, such as the missile shield and the fans, are supported at the distal ends of three relatively long lift rods, resulting in an unstable structure that may subject the lift rods to undesirable compressive, bending, and torsional stresses. Malandra et al. also does not provide a structure for putting a shroud around the CRDMs. In U.S. Pat. No. 4,830,814, Altman discloses an integrated head package having a missile shield that is slidably mounted near the distal end of three lift rods connecting to the reactor vessel closure head lifting lugs. A shroud is shown disposed about the CRDMs. Similar to the apparatus disclosed by Malandra et al., however, the heavy missile shield and lifting rig are installed at the distal end of three elongate lift rods that are connected at their proximal end to the reactor vessel closure head lifting lugs. The Altman apparatus, therefore, will also produce relatively high local loads in the reactor vessel lifting lugs and head. Altman also does not disclose any system for cooling the CRDMs. In U.S. Pat. Nos. 6,546,066 and 6,618,460, which are hereby incorporated in their entirety, the present inventor discloses an improved integrated head assembly having a cylindrical shroud on a ring support mounted to the reactor vessel closure head. The disclosed integrated head assembly includes a forced air cooling system, missile shield, and seismic support system. The integrated head assembly is removable in a single lift. In some applications, however, it may be preferable to have at least portions of the head assembly supported separately from the reactor head, and/or to divide the lift for removing these systems into more than a single lift. There remains a need, therefore, for an integrated head assembly for a pressurized water reactor that can be removed from the reactor vessel integrally with the reactor vessel closure head and that does not introduce undue local stresses at the reactor vessel closure head and lifting lugs. A two-part integrated head assembly (“IHA”) for a commercial nuclear reactor is disclosed. The two-part IHA relieves the loads on the reactor vessel, as compared with prior one-piece integrated head assemblies. For example, and not by way of limitation, it will be appreciated that the two-part integrated head assembly is particularly beneficial in reactor systems that do not incorporate load-transferring seismic tie rods on their service structures at the refueling floor elevation. A lower portion of the IHA attaches to the reactor vessel closure head, for example utilizing a ring beam that distributes the loads and lift rods that attach to lifting lugs on the reactor vessel closure head. The lower portion may include a shroud or outer wall the surrounds the control element drive mechanisms, and that define vertical air flow channels that fluidly connect to an annular plenum. A baffle system may also be provided to more predictably control the cooling air flow path. The upper portion includes vertical support beams, and horizontal support beams. The horizontal support beams engage auxiliary structure in containment, such as the steam generator walls, to support the upper portion of the IHA during use. When access to the interior of the reactor vessel is required, for example during refueling outages, the upper portion can be removed and moved to an alternate location for storage, and may be supported by the vertical support beams. In a preferred embodiment, the lower portion is removable as a unit with the reactor vessel closure head by disengaging the reactor vessel attachment hardware, and lifting the lift rods. Typically, the reactor vessel closure head and lower portion of the IHA may then be stored in a location typically reserved for the closure head. In an embodiment of the IHA, the duct has a releasable and flexible joint for attachment to the annular plenum on the lower portion. Platforms and other access assemblies may also be provided on the upper portion. It is also contemplated that heat exchange equipment such as a chiller may be provided, most conveniently in the fan plenum, to cool the cooling air prior to expulsion into containment. In an embodiment of the IHA, the upper portion includes a missile shield that is adapted to stop or hinder the expulsion of control elements and/or CEDMs during certain accident scenarios. In a typical light water pressurized water reactor (“PWR”) design for commercial power generation, a number of individual components are assembled and located over the reactor vessel closure head inside the containment structure. These components generally must be disassembled for reactor refueling and then reassembled during every refueling outage. The disassembly and assembly procedures require a considerable amount of time and, in particular, require significant worker time inside containment. As a result, workers may receive significant radiation dosage. To minimize the critical path time and radiation dosage during refueling outages, an integrated head assembly (“IHA”) has been designed by the present inventor, and disclosed in U.S. Pat. Nos. 6,546,066 and 6,618,460, which are incorporated herein by reference. The one-piece IHA is assembled into a single unit and provides in a single mechanical assembly most or all primary head area components. The one-piece IHA is attached to the reactor vessel closure head, and permits access to the vessel head attachment hardware, permitting all of these components to be moved in a single lift. The IHA also provides a forced air convection system that improves the efficiency of the control rod drive mechanism/control element drive mechanism cooling (“CEDM”). The IHA saves a significant amount of critical path time and radiation dosage during refueling outages. However, in some instances, integrating the various head area components, and including a more efficient CEDM cooling system, results in an IHA total weight that is greater than the total weight of the original head area components supported by the reactor vessel head in the original design. In some PWR designs—for example, in certain Westinghouse and Babcock & Wilcox PWRs—the weight of the IHA is supported in part by the reactor vessel and in part by the containment cavity walls (e.g., through seismic support structures). In other PWR designs—for example, in certain PWRs designed by Combustion Engineering—all of the weight of the single-piece IHA would be supported by the reactor vessel alone. It will be appreciated by persons of skill in the art that any additional weight from an IHA on the reactor vessel and the containment walls requires additional evaluation of the reactor vessel, reactor coolant loop, and the containment walls. Generally, the load capacities of the concrete walls are significantly greater than the loads applied by the weight of the IHA on the concrete walls. A reevaluation of the loads on the reactor vessel and the reactor coolant loop is necessary to assure that the system meets all requirements in the commercial nuclear power industry. A two-part IHA has been designed, as disclosed herein, that does not require the reactor vessel to support the entire weight of the two-part IHA. The two-part IHA disclosed herein does not require that the reactor vessel support significantly more weight than is currently supported in a typical conventional installation without an IHA, and preferably does not require that the reactor vessel support any more than the original head design load. In addition, in the two-part IHA disclosed herein, the two parts of the IHA can be removed sequentially, reducing the maximum weight that must be moved in any single lift. It will be appreciated by those of skill in the art that the present two-part IHA design may therefore eliminate any need for a polar crane upgrade. Refer now to FIGS. 1A, 1B, 1C and 1D showing a currently-preferred embodiment of a two-part IHA 100 in accordance with the present invention. FIG. 1A shows the IHA 100 installed on a reactor vessel closure head 90; FIG. 1B shows a more detailed drawing of the assembled IHA 100; FIG. 1C shows the IHA 100 removed from the reactor vessel (not visible), with the upper portion 160 stored separate from the lower portion 110. A side view of the assembled IHA 100 is shown in FIG. 1D. The two-part IHA 100 provides a shroud providing radiation shielding, and includes a forced air convective cooling system that directs air over the CEDMs to provide cooling as discussed in detail below, and a missile shield system for protecting against potential ejection of elements from the reactor in certain accident scenarios, also discussed below. The two-part IHA 100 includes a generally cylindrical lower portion 110, having an outer wall portion that is disposed about, and shrouds the CEDMs 94. An upper portion 160 of the IHA 100 provides cooling components such as the fans 166, and a missile shield 169 (FIG. 3B), and includes an opening 161 (FIG. 1C) at its bottom end that is sized to accommodate the top end of the lower portion 110. A duct 150 extends between the lower portion 110 and the upper portion 160. The perimeter of the lower portion 110 is sized such that the lower portion 110 does not interfere with regular access to the attachment hardware for the reactor vessel closure head 90. The weight of the upper portion 160 is preferably supported by concrete walls 92—for example, the steam generator walls (shown in phantom in FIG. 1A), or other suitable structure in the containment area around the reactor vessel. Persons of skill in the art will readily recognize that such concrete walls 92 are typically already present in the reactor containment, e.g., in some typical installations a concrete missile shield is supported by the concrete walls 92. The lower portion 110 is attached to and supported by the reactor vessel closure head 90. The lower portion 110 is substantially mechanically uncoupled from the upper portion 160—that is, the lower portion 110 does not support any substantial fraction of the weight of the upper portion 160 and the upper portion 160 does not support any substantial fraction of the weight of the lower portion 110. When the lower portion 110 and upper portion 160 are assembled and in use, they cooperatively provide efficient cooling for the CEDMs 94 (indicated schematically in FIG. 1A), and other typical IHA functions. As shown in FIG. 1C, during outages that require the reactor vessel closure head 90 to be removed, the upper portion 160 of the IHA 100 is removed first and placed in a suitable location 96, typically within containment. The lower portion 110 of the IHA 100 is then removed, preferably with, and while still attached to, the reactor vessel closure head 90, and placed in a vessel head parking location 98. FIG. 2A shows a perspective view of the lower portion 110 of the IHA 100 mounted on a reactor vessel closure head 90 with the CEDMs 94 extending upwardly, generally shrouded by the lower portion 110. An array of messenger wires 115 are shown disposed at the top of the lower portion 110, although it will be appreciated that messenger cables 115 may alternatively, or additionally, be provided on the upper portion 160. The lower portion 110 includes a ring beam 112 having a plurality of integral feet or saddle members 114 that approximately conform to the shape of the closure head 90. The ring beam 112 sits atop the closure head 90 and the saddle members 114 distribute the weight load on the closure head 90. The ring beam 112 engages lifting lugs 95 (one visible in FIG. 2A) formed in the closure head 90. A plurality of lifting rods 116 also engages the lifting lugs 95 and extends upwardly, generally terminating in a connector 117, the connectors 117 being adapted to engage a tripod (not shown) to facilitate lifting of the IHA lower portion 110 and closure head 90. A suitable ring beam is disclosed in U.S. Pat. No. 6,618,460 (incorporated by reference). In some embodiments of the present invention, particularly suitable for installations wherein the reactor vessel closure head 90 does not have lift lugs, the lifting rods 116 may attach directly to the ring beam 112. In the preferred embodiment, the IHA lower portion 110 includes a bottom segment 120, a middle segment 130, and a top segment 140. The bottom segment 120 is supported by the ring beam 112. The middle segment 130 is attached to the bottom segment 120 and the top segment 140 is attached to the middle segment 130. The segments 120, 130, and 140 cooperatively define a generally cylindrical structure that shrouds the CEDMs 94 and is small enough in diameter such that it does not interfere with tightening of the head bolts (not shown) on the top of the closure head 90. Although the three-segment construction of the IHA lower portion 110 is currently preferred, it is contemplated that the lower portion may alternatively be constructed in one segment, two segments, or more than three segments, without departing from the present disclosure. FIG. 2B shows the bottom segment 120 of the IHA lower portion 110, attached to the reactor vessel closure head 90 (shown in phantom), with the ring beam 112. The bottom segment 120 includes an outer wall 122 and two inner walls 124, each inner wall 124 defining cooperatively with the outer wall 122 a substantially vertical channel 126 that is open at the top and bottom. A plate 128 having a plurality of apertures 129 sized and positioned to accommodate the CEDMs 94 extends horizontally across the bottom segment 120 at an intermediate elevation. FIG. 2C shows the middle segment 130 in isolation. The middle segment 130 includes an outer wall 132 and two inner walls 134, each inner wall 134 defining cooperatively with the outer wall 132 a substantially vertical channel 136 that is sized and positioned to substantially align with and continue the channels 126 in the bottom segment 120. The middle segment includes one or more air inlet ports 131 that may include a closable door (not shown) for providing air flow into the lower portion 110. FIG. 2D shows the top segment 140 in isolation. The top segment 140 includes an outer wall 142 and an inner wall 144. The outer wall 142 and inner wall 144 are closed at the top 145 and partially closed at the bottom 146, the bottom 146 having downwardly facing ports or openings 147 (one visible in FIG. 2D) that are sized and positioned to match and continue the channel 136 in the middle section 130. The top segment 140, therefore defines a substantially annular plenum volume for the channels 136. An air outlet aperture 149 is also provided in the outer wall 142 that is adapted to engage the duct 150 to fluidly connect to the fan plenum 165 (see FIG. 1A), as discussed below. Refer now to FIGS. 3A and 3B that show the upper portion 160 of the integrated head assembly 100 in isolation. The upper portion 160 includes four upright supports 162 having foot pads 164 at a lower end and lifting connectors 163 at an upper end. The lifting connectors 163 are preferably adapted to engage a tripod (not shown) to facilitate lifting of the IHA upper portion 160, to temporarily relocate the upper portion 160, for example during refueling outages. The upright supports 162 engage a pair of transverse beams 170 that are adapted to extend between the concrete walls 92 (see FIG. 1A). The beams 170 are engineered to support the weight of the upper portion 160. As seen in FIG. 1C, the upright supports 162 are designed such that during outages the upper portion 160 can be set on a flat surface, supported by the upright supports 162. Optionally, horizontal plates 172 may be provided extending between the transverse beams 170, to define a platform, and ancillary access structures 173, to facilitate access to the upper portion 160. The upper portion 160 further comprises a fan plenum 165 fluidly connecting a plurality of fans 166 (four shown) that are oriented to draw air upwardly through the plenum 165. The fan plenum 165 is connectable via the duct 150 to the plenum defined in the top segment 140 of the lower portion 110, as discussed above. The upper portion 160 may further include cable supports and bridges 180, and additional work platform(s) 182. In a preferred embodiment, a missile shield 169 is incorporated into, or attached to, the fan plenum 165, wherein the missile shield 169 is defined to be a structure engineered to hinder or protect against the potential ejection of control elements, control element drive mechanisms, or the like in certain over-pressure accident scenarios. In a preferred embodiment, the duct 165 is releasably attached to the air outlet aperture 149 in the lower portion 110 of the IHA 100 by means of a flexible joint so that no significant load from the duct 165 will be transmitted to the lower portion 110 of the IHA 110. The cooling air flow path for the currently preferred integrated head assembly 100 can now be appreciated. The fans 166 draw air through the fan plenum 165, through the duct 150, through the channels defined by the bottom segment 120, middle segment 130 and top segment 140 of the lower portion 110, and in through the air inlet port 131, such that the air flows about the CEDMs to remove excess heat prior to being drawn to the fan plenum 165. It is contemplated that the fan plenum 165 may include chillers 168 (FIG. 3B) or other supplemental heat exchangers to cool the circulated air prior to expelling the air back into the containment building. In one contemplated embodiment the chillers 168 are plumbed to an external cold water source (not shown). It will also be appreciated that the lower portion 110 may include baffles to more efficiently direct the air about the CEDMs. A suitable baffle structure is disclosed in the incorporated U.S. Pat. No. 6,618,460. Some advantages of the two-part IHA 100 disclosed herein are as follows: 1. The lower portion 110 of the two-part IHA 100 provides features that are provided by prior IHAs. For example, the shrouding function of the lower portion 110 provides radiation protection. The lower portion may include inspection doors, and/or means for providing easy access to the dome insulation; easy access to the CEDM nozzles; and/or easy access to other reactor vessel systems such as the reactor vessel level indicator system (“RVLIS”), the core exit thermocouple system (“CET”), the reactor vessel head vent system (“RVHVS”), and the heated junction thermocouple system (“HJTC”). The two-part IHA 100 provides a modified and improved CEDM cooling system. It is contemplated that the two-part IHA may incorporate CEDM seismic support and tie rods, in particular for certain Westinghouse and B & W-designed PWRs. 2. The design of the lower portion 110 of the two-part IHA 100 may preferably be optimized such that the total weight of the lower portion is very close to the total weight of the originally designed reactor vessel head service structure. This will eliminate or reduce the need for any reevaluation of the reactor vessel and reactor coolant loop for added weight. 3. The upper portion 160 of the two-part IHA 100 provides features that are provided by prior one-part IHAs such as the missile shield assembly, and cooling equipment. 4. In the present design, an existing concrete missile shield in current containment structures may be removed or eliminated when using the present invention. The concrete missile shield generally weighs at least a few hundred thousand pounds and is typically supported by a concrete structure, for example steam generator walls. The total weight of the upper portion 160 of the two-part IHA 100 may be one-fourth (¼) to one-third (⅓) the weight of an existing concrete missile shield. Therefore, the concrete structure that is supporting the upper portion 160 of the IHA 100 will generally not require any structural modifications to take the loads from the upper portion 160 of the two-part IHA 100. 5. During refueling outage, the upper portion 160 of the two-part IHA 100 may be stored at the same location that the concrete missile shield is typically stored in current facilities. The lower portion 110 of the IHA 100 may be stored on the reactor vessel head stand. Therefore, the two-part IHA 100 is suitable for use in existing reactor installations, without requiring substantial modifications of the containment structure. 6. Since the lower portion 110 of the IHA 100 is intended to weigh approximately the same as the original head service structure, it is possible to reuse the existing lift rod and the tripod assemblies when the present invention is implemented as a modification or upgrade to an existing facility. 7. The upper portion 160 of the IHA 100 provides additional walkway(s) for workers to walk above the CEDMs in the middle of the lower portion 110 of the IHA 100. It also provides additional safety for workers while working on the CEDM inside containment. 8. The work platform(s) in the upper portion 160 of the IHA 100 may be located to provide easy access to disconnect the duct 150 from the lower portion 110 of the IHA 100. 9. The upper portion 160 of the IHA 100 provides an upper air plenum 165 that is large enough to provide one or more water chiller(s) 168 therein to cool the exhaust air from the CEDM cooling system prior to being ejected by cooling fans 166 into containment. 10. The upper air plenum 165 in the upper portion 160 of the IHA 100 provides additional space that may productively be used, for example allowing for the use of four cooling fans 166 compared to the prior art IHA designs that utilize only three cooling fans. Other advantages and benefits of the two-part IHA 100 disclosed herein will be apparent to persons of skill in the art based on the present disclosure, and on the disclosure in patents that are incorporated herein by reference. While the preferred embodiment of the invention has been illustrated and described, it will be appreciated that various changes can be made therein without departing from the spirit and scope of the invention.
	claims	1. A method for operating a pressurized water nuclear reactor comprising a core containing nuclear fuel assemblies comprising nuclear fuel rods, the method comprising:operating the nuclear reactor during successive cycles with between each cycle, steps for replacing spent nuclear fuel assemblies with fresh nuclear fuel assemblies, including:operating the reactor for at least one plutonium-equilibrium cycle during which the core contains plutonium-equilibrium nuclear fuel assemblies, the plutonium-equilibrium nuclear fuel assemblies comprising, before irradiation, nuclear fuel rods exclusively based on uranium and plutonium mixed oxide, the nuclear fuel rods of each plutonium-equilibrium nuclear fuel assembly having a same isotope composition of nuclear fuel and a same nominal total plutonium mass content; thenoperating the reactor for transition cycles, at least some of the plutonium-equilibrium nuclear fuel assemblies being progressively replaced, during the replacement steps preceding transition cycles, with zoned transition nuclear fuel assemblies and then uranium-equilibrium nuclear fuel assemblies, the zoned transition nuclear fuel assemblies each comprising a central zone comprising nuclear fuel rods containing, before irradiation, uranium oxide and not containing any plutonium oxide, and a peripheral zone extending along outer faces of the zoned transition nuclear fuel assembly, the peripheral zone only comprising, before irradiation, nuclear fuel rods exclusively based on uranium and plutonium mixed oxide, the uranium-equilibrium nuclear fuel assemblies only comprising, before irradiation, nuclear fuel rods containing uranium oxide and not containing any plutonium oxide; and thenoperating the reactor for at least one uranium-equilibrium cycle in which the core contains uranium-equilibrium nuclear fuel assemblies, the uranium-equilibrium nuclear fuel assemblies only comprising, before irradiation, nuclear fuel rods containing uranium oxide and not containing any plutonium oxide. 2. The method as recited in claim 1 wherein during the uranium-equilibrium cycle, the core only contains uranium-equilibrium nuclear fuel assemblies comprising only, before irradiation, nuclear fuel rods containing uranium oxide and not containing any plutonium oxide. 3. The method as recited in claim 1 wherein, during the plutonium-equilibrium cycle, the core only contains plutonium-equilibrium nuclear fuel assemblies. 4. The method as recited in claim 1 wherein, during the plutonium-equilibrium cycle, the plutonium-equilibrium nuclear fuel assemblies only comprise before irradiation nuclear fuel rods exclusively based on uranium and plutonium mixed oxide. 5. The method as recited in claim 1 wherein, during the plutonium-equilibrium cycle, the nuclear fuel rods of all the plutonium-equilibrium nuclear fuel assemblies have a same isotope composition of nuclear fuel and a same nominal total plutonium mass content. 6. The method as recited in claim 1 wherein at least some of the zoned transition nuclear fuel assemblies each comprise in the central zone thereof poisoned nuclear fuel rods, the poisoned nuclear fuel rods containing before irradiation, at least one consumable neutron poison. 7. The method as recited in claim 1 wherein, in at least some of the zoned transition nuclear fuel assemblies, the nuclear fuel rods of the peripheral zones have nominal plutonium fissile isotope contents of less than those of nuclear fuel rods of plutonium-equilibrium nuclear fuel assemblies. 8. The method as recited in claim 1 wherein during the replacement step preceding a first transition cycle, first zoned transition nuclear fuel assemblies are loaded into the core, wherein during the replacement step preceding a second transition cycle, second zoned transition nuclear fuel assemblies having central zones including nuclear fuel rods having, except for possible poisoned nuclear fuel rods, uranium 235 enrichments different from those of the nuclear fuel rods of the central zones of the first zoned transition nuclear fuel assemblies, are loaded into the core. 9. The method as recited in claim 8 wherein, except for the possible poisoned nuclear fuel rods, the nuclear fuel rods of the central zones of the second zoned transition nuclear fuel assemblies have substantially a same uranium 235 enrichment as the nuclear fuel rods of the uranium-equilibrium nuclear fuel assemblies. 10. The method as recited in claim 1 wherein the zoned transition nuclear fuel assemblies are not loaded into an outer peripheral layer of the core and at least some of the zoned transition nuclear fuel assemblies are loaded in a layer immediately adjacent to the outer peripheral layer of the core.
	claims	1. A process for irradiating products by means of a high energy X-ray beam source in an installation having an irradiation chamber, said process comprising the steps of:determining the density of the products to be irradiated,in order to irradiate said products as a stack, predetermining, on the basis of said density, the optimal size of the product stack able to optimize the throughput of the installation and/or the dose uniformity ratio (DUR),in the irradiation chamber, loading products as a stack of said optimal size onto rotation means located in front of the X-ray beam source,while rotating the rotation means around a rotation axis, irradiating said products from a lateral side of said product stack without using a collimator wherein said steps are controlled by a controlling means. 2. The process according to claim 1, wherein the rotation speed of the rotation means in front of the X-ray beam source is maintained constant. 3. The process according to claim 2, wherein the rotation speed of the rotation means in front of the radiation beam source is maintained constant by the action of the controlling means at a value depending upon predefined parameters. 4. The process according to claim 1, wherein the products are carried on pallets and the stack is formed by at least two contiguous pallets. 5. The process according to claim 4 wherein the stack comprises at least four product pallets. 6. The process according to claim 1 wherein the pallets in the stack are in a plane perpendicular to the rotation axis of the rotation means. 7. The process according to claim 5 wherein the four product pallets to be irradiated are rectangular product pallets and form together a square base with an open column at the centre of the square base. 8. The process according to claim 7, wherein the centre of the square coincides with the rotation axis of the rotation means. 9. The process according to claim 5 wherein the four product pallets to be irradiated are rectangular pallets having each at least one corner, and said four product pallets are arranged in such a way that said corner of each pallet coincides in a contact point with one corner of the other three pallets. 10. The process according to claim 9 wherein the contact point is located on the rotation axis of the rotation means. 11. The process according to claim 1, adapted for irradiating products under bulk form or under the form of small parcels, wherein the product stack is maintained in at least one cylindrical container having an internal volume. 12. The process according to claim 11, wherein said products are arranged in said cylindrical container so as to fill the total internal volume of said cylindrical container. 13. The process according to claim 11, wherein said products are arranged in said cylindrical container so as to let an open column along the center axis of the cylindrical container. 14. The process according to claim 11, wherein said cylindrical container is selected from a set of cylindrical containers consisting of mortars and cylindrical baskets, said cylindrical containers having a diameter near to said determined optimal size. 15. The process according to claim 1, wherein the irradiation of the products is performed by batches of products of similar densities. 16. An apparatus for irradiating products, said apparatus comprising:a high energy X-ray beam source, for irradiating the products from a lateral side with a beam directed along a first direction substantially perpendicular to said lateral side, and scanned along a second direction substantially perpendicular to said first direction,an irradiation chamber, where irradiation of the products can be performed, said irradiation chamber comprising rotation means for rotating said products around a rotation axis parallel to said second direction,said rotating taking place in front of said X-ray beam source at a constant rotation speed, during irradiation, said rotation means comprising means for receiving the products;wherein said apparatus does not comprise a collimator, and said means for receiving the products are adapted to receive products loaded thereon as a stack the size of which is variable depending on the density of said product,wherein said apparatus is adapted to receive a product stack comprising products carried on pallets,wherein the product stack comprises at least four contiguous pallets,wherein that at least four product pallets to be irradiated are rectangular pallets having each at least one corner and said four product pallets are arranged in such a way that said corner of each pallet coincides in a contact point with one corner of the other three pallets,and wherein the contact point is located on the rotation axis of the rotation means. 17. An apparatus for irradiating products, said apparatus comprising:a high energy X-ray beam source, for irradiating the products from a lateral side with a beam directed along a first direction substantially perpendicular to said lateral side, and scanned along a second direction substantially perpendicular to said first direction,an irradiation chamber, where irradiation of the products can be performed, said irradiation chamber comprising rotation means for rotating said products around a rotation axis parallel to said second direction,said rotating taking place in front of said X-ray beam source at a constant rotation speed, during irradiation, said rotation means comprising means for receiving the products;wherein said apparatus does not comprise a collimator, and said means for receiving the products are adapted to receive products loaded thereon as a stack the size of which is variable depending on the density of said products,wherein said apparatus is adapted to receive a product stack comprising products carried on one or more pallets, wherein each of said one or more product pallets to be irradiated comprises one corner and a contact point and said contact point is located on the rotation axis of the rotation means and the corner and contact point of said one or more product pallets coincides.
042773083	description	DETAILED DESCRIPTION According to the drawing, there is shown a reactor flux level 8 which is increasing monitonically with time and which is monitored by a monitoring circuit. The monitoring circuit includes a proportional counter 10 which is a device which develops an output pulse in response to a neutron being incident thereon, i.e., it is a pulse-type neutron detector. The proportional counter 10 is positioned in the nuclear reactor environment and the output pulses therefrom are applied to a pulse amplifier system 12 and then to a comparator 14. Comparator 14 separates neutron-induced pulses for unwanted pulses on the basis of pulse amplitude. Thus, comparator 14 will output neutron-induced pulses which are applied to counter 16. Counter 16 counts the pulses it receives. The pulses from comparator 14 are also applied to a programmable divider 18 which generates one output pulse after each n pulses from the comparator 14, thus, in effect dividing by n the output of comparator 14. Programmable divider 18 could be set to any integer value as will be described. The output of divider 18 is applied to counter 20 which counts the number of pulses from divider 18. Programmable timer 22 is coupled to counter 16 and counter 20 and develops a command signal every .DELTA.t time period with .DELTA.t being determined as will be described. In response to a command signal from timer 22, counter 16 updates the value contained in memory register 24. Also in response to the command signal from timer 22, i.e., after each time period .DELTA.t, each counter 16 and 20 is reset to zero so that they begin a new counting cycle. Comparator 26 compares the value of the output of counter 20 with value stored in register 24. If the value of counter 20 is greater than or equal to the value stored in register 24, comparator 26 generates a trip signal. Scram generator 28 is responsive to a trip signal from comparator 26 to generate a reactor scram signal, indicating that the reactor power level is increasing at an unacceptable rate. The count-factor-increase time monitor operates as follows: beginning at zero counts, counter 16 counts the number of neutron-induced pulses from detector 10 for the first .DELTA.t time period. At the end of the first .DELTA.t time period, the count achieved is stored in register 24. During the next or second .DELTA.t time period, counter 16 again starts at zero and counts the number of neutron-induced pulses from detector 10. Also during the second .DELTA.t time period, counter 20 is counting the number of neutron-induced pulses from detector 10 divided by n. During the second .DELTA.t time period, comparator 16 is constantly comparing the count of counter 20 with the value stored in register 24, which will be the value counted by counter 16 during the first .DELTA.t time period, and if the count of counter 20 is at any time equal to or greater than the value stored in memory register 24, a trip signal is generated at that time by comparator 26. This indicates that the rate of change of the reactor power level has exceeded the allowed value and the reactor should be scrammed. At the end of each .DELTA.t time period, the value stored in register 24 is updated to the value counted by counter 16, and the counters 16 and 20 are zeroed. A new monitoring cycle is begun. It is apparent that during the first cycle of operation, register 24 will have no value stored therein. To start monitoring, one can use a preset input count 30 to preset register 24, so that this preset value is initially set in register 24. The comparators can be inhibited during this preset operation. The theory underlying the operation of the monitor is that the reactor period and the count-factor-increase have been determined to obey the following relation, providing one assumes that reactor period is constant for a given time span equal to twice .DELTA.t or longer: ##EQU1## where T.sub.j-(j+1) is an asymptotic period measured between time intervals t.sub.(j-1) and t.sub.(j+1), N.sub.(j+1) is the number of counts occurring between t.sub.j and t.sub.(j+1), PA1 N.sub.j is the number of counts occurring between t.sub.(j-1) and t.sub.j, and PA1 .DELTA.t.sub.j-(j+1) is the time duration between time t.sub.j and time t.sub.(j+1). It is assumed that .DELTA.t.sub.(j-1)-j =.DELTA.t.sub.j-(j+1) =.DELTA.t.sub.(j+1)-(j+2), etc. Consider the case where N.sub.(j+1) /N.sub.j =2. Programmable divider 18 will then have n=2 and the monitor will observe count-doubling time. Fundamentally, the count-doubling time measuring circuit measures the time necessary for the number of counts in the interval from t.sub.j to t.sub.(j+1) to equal twice the number of counts in the interval from t.sub.(j-1) to t.sub.j. Thus it is generally more meaningful to think of this circuit in terms of count-doubling time rather than period. If this time is less than a prescribed value a scram is initiated as soon as N.sub.(j+1) /N.sub.j is equal to 2. The scram will occur on the basis of the time required for N.sub.(j+1) /N.sub.j to equal 2 regardless of whether the reactor is on an asymptotic or transient period. In practice, if the times required for N.sub.(j+1) to be twice N.sub.j are greater than the prescribed value, the circuits are reset at definite .DELTA.t intervals and N.sub.(j+1) replaces N.sub.j as the reference count; no scram is initiated. The ratio N.sub.(j+1) /N.sub.j may, of course, be any appropriate constant. However, the factor 2 is certainly acceptable from safety considerations for low and medium power level reactor operation and is convenient from circuit considerations. The statistics associated with N.sub.(j+1) /N.sub.j =2 are also acceptable. If, for example, a scram is desired when N.sub.(j+1) /N.sub.j =2 occurs in less than 4 seconds, a counting rate of 100 counts per second would result in the order of 400 counts during this interval. The standard deviation would be about 20 and a 10.sigma. deviation would be only about 50% of the total count required to effect a scram. This fluctuation, although very large and very improbable, will not result in a scram if superimposed upon any reasonable operating period. With n=2 and .DELTA.t selected at 5 seconds, i.e. timer 22 set at 5 seconds, a reactor scram will be developed with T.apprxeq.7.2 seconds. Generally one selects T in the range of 5-10 seconds. The minimum .DELTA.t that can be used is a statistical limitation in that there must be a sufficient .DELTA.t to allow for meaningful counts. The count-factor-increase time monitoring circuit described is not by itself a good source of information for display or recording purposes. One way to obtain visual display information is to use separate comparator circuitry to determine the length of time for N.sub.(j+1) /N.sub.j to reach some value such as 1.10. This time is a measure of the rate of the increase of the reactor power level. The time can be recorded by a scaler and transmitted to the display devices. Another method simply uses a microcomputer or programmable calculator 32 to interrogate counter 16 and counter 20 for information at appropriate time intervals, and then calculate the period or count-doubling times. This new circuit is simple, fast and is capable of assuring that the reactor power level cannot rise too rapidly. By selecting a suitable N.sub.2 /N.sub.1 ratio, the probability for spurious scrams can be negligible without compromising reactor safety. An inexpensive microcomputer or programmable calculator can be operated in parallel with the safety circuit to provide frequent updating of period or rate of change information for display and recording purposes.
061223395	summary	BACKGROUND OF THE INVENTION Field of the Invention The invention relates to a method of operating a boiling-water reactor which is an unstable state as a result of local oscillation of a physical variable (in particular of the power or of the neutron flux associated therewith). In addition, the invention relates to a device for carrying out this method and to a method and a device for monitoring the unstable reactor state. The nuclear fission determining the power of a nuclear reactor is controlled by moving absorber elements into the reactor core in order to attenuate the neutron flux. In this arrangement, measuring lances having sensors for the flux of thermal neutrons are distributed over the reactor core, in order to register the current state. In order to adjust a desired operating state, it is also necessary for the throughput of coolant (cooling water), which serves at the same time as a moderator, to be adapted to the respective state. The coolant enters in liquid phase into the reactor core from below, flows through the fuel elements, in which it partially evaporates, and emerges from the core as a vapor phase/liquid phase mixture, as a result of which the fuel/moderator ratio in the various parts of the fuel elements is changed. At the same time, however, the flow conditions are changed, in particular the location at which the single-phase flow, with which the liquid coolant enters the fuel elements, changes into the two-phase flow of the liquid/vapor mixture. In this case, at high power and low coolant throughput, unstable conditions have been observed in which this phase boundary goes into an oscillating motion, which results in a pulsation of the moderator density and the power, which has a bearing on the cooling capacity and the movement of the phase boundary. In this case, periodic temperature fluctuations with considerable peak values may occur in the fuel elements. The permissible power maximum of the fuel elements is mainly limited by the temperature resistance of the materials used in the fuel elements. If an upper temperature limit is exceeded, the materials loose their mechanical, chemical and physical properties and can undergo irreversible changes, which can force an exchange of the fuel elements. Therefore, care must be taken that this thermal-hydraulic upper power threshold (and hence a thermal-hydraulic threshold value A.sub.th, of the neutron flux) in the reactor is not exceeded. Safety provisions in the reactor operation therefore call for a rapid shutdown of the reactor (so-called "SCRAM"), in the event the threshold value is exceeded. In such an emergency program, all the control rods are rapidly moved in and the corresponding cooling capacity is set. Following such a SCRAM, the reactor is restarted according to a predetermined startup program, so that there is a considerable disturbance to the reactor operation. In addition, the fuel elements have to be changed for safety reasons, if the thermal-hydraulic threshold value has been reached many times or over a relatively long period of time. The art is therefore concerned with detecting and damping an unstable state of this type as early as possible, before the power pulsations reach the vicinity of the thermal-hydraulic threshold value. It has been shown that these pulsations always occur in a frequency range between about 0.3 and 0.7 Hz and have a very constant frequency. The method described in U.S. Pat. No. 5,174,946 to Watford et al. (=EP 0 496 551) for monitoring the power fluctuation band for nuclear reactors is based on that fact. That process utilizes the flux as a measured variable for the unstable state caused by the local oscillation of a physical variable, the measuring lances mentioned ("local power range monitor-strings", LPRM strings) being used for this flux measurement. Each such lance normally contains four sensors, whose signals are observed anyway for power control purposes, then further processed and documented. Each of these four sensors in each measuring lance is used, two sensors being assigned to a first monitoring system, the two remaining sensors being assigned to a redundant second monitoring system. Each monitoring system thereby contains two monitoring channels, each sensor signal of a measuring lance being assigned to a different monitoring channel. Different subdivisions of the reactor into individual regions ("monitoring cells") are in this case based on the two monitoring channels of a system, each cell being bounded by four measuring lances in order to form a corresponding region signal. Depending on the location of the measuring lance in the core (in the interior of the core or at the edge of the core), a sensor signal in each monitoring channel belongs to two, three or four cells. As a result of this multiple use of the sensor signals, it is intended to achieve the situation where virtually the state of each individual fuel element can be monitored and identified by means of the influence which it has on the sensor signals of the individual cells. To this end, provision is made that an alarm is set in a system only when both monitoring channels respond. Although it is sufficient for the alarm to be given by one of the two systems, only simple redundancy is provided thereby. A further disadvantage is that virtually all the monitoring channels are affected by an erroneous measurement or a complete failure of a measuring lance, it being possible in the case of an edge position of the measuring lance, for example, that simultaneously a plurality of cells are no longer being monitored properly. The state of the individual cells (regions) is monitored by initially monitoring in a plausibility control whether the individual sensor signal exceeds a specific lower threshold value and is operating properly. In the case of a sensor defect, the signals belonging to this cell are not evaluated further. By means of summing all the sensor signals of a region, a current region signal is formed which is suppressed, however, if (for example as a result of an erroneous measurement) a plausibility monitoring yields the fact that the region signal does not achieve a predefined minimum value. The region signal is then filtered and related to an average over time, the time constant of which is greater than a period of the oscillation, so that a relative current region signal is produced which indicates by how many percent the current power of the region lies above or below the average. If this current value exceeds a power limit (for example 120%), a check is then made as to whether this is a once-off transition state (so-called "transient") which for example constitutes only an aperiodic transition to a new operating state predefined by the control, without exciting an oscillation. In this case, this is not therefore a critical oscillation in the frequency band from 0.3 to 0.7 Hz, so that no intervention is carried out as long as a threshold value A.sub.max, lying in the vicinity of the thermal-hydraulic threshold value A.sub.th, is not reached. In order to detect the critical oscillation, instead an examination is made to see whether, in a time interval corresponding to this critical frequency band, the value does not also fall below a corresponding threshold value (e.g. 80%) following the exceeding of a limiting value A.sub.o as is necessary for an oscillation. If it is determined in this way that--corresponding to an oscillation--a lower extreme value follows an upper extreme value of the flux, a check is further made as to whether another upper extreme value follows this lower extreme value, and whether this following upper extreme value exceeds an alarm value which lies above the extreme value detected first by a predefined factor (e.g. 1.3). If this is so, then after this one oscillation period it is already concluded that there is a growing, i.e., increasing oscillation, in which the exceeding of A.sub.th is threatened, and the SCRAM is initiated even before the value A.sub.max is reached. With an eye to the present invention, reference is made at this point that, although the above-described prior art monitors whether the oscillation is growing at a rate lying above the predefined factor (here 1.3), the growth (rate of increase) of the extreme values is not itself measured. This factor (1.3) is also relative in as much as it is related to the extreme value detected first, but it independent of the rate of increase. In addition, reference is made to the fact that although it is checked whether the time interval between the detected extreme values corresponds to the critical frequency band of 0.3 and 0.7 Hz, no check is made as to whether the next extreme value A.sub.n+1, follows in practice at the same interval DT.sub.n, which is given by the previously detected upper extreme value (denoted A.sub.n-1, point in time T.sub.n-1) and the presently detected lower extreme value (A.sub.n, point in time T.sub.n), after this point in time T.sub.n. Those skilled in the art of reactor control and monitoring will appreciate that the usual techniques for the monitoring and documentation of the sensor signals apply and they will therefore readily be able not only to register the extreme values A.sub.n-1, A.sub.n, A.sub.n+1 . . . but also the points in time T.sub.n-1, T.sub.n, T.sub.n+1 . . . at which these extreme values occur. The person in charge of monitoring could therefore readily suppress the corresponding region signal if the time interval DT.sub.n =T.sub.n -T.sub.n-1 deviates significantly (for example 0.1 seconds) from the time interval DT.sub.n+1 =T.sub.n+1 -T.sub.n. However, U.S. Pat. No. 5,174,946 contains no advice on this point. In the state of that prior art, therefore, no attention is initially paid to an oscillation whose (unmeasured) rate of increase lies below the set factor (1.3); rather, intervention is considered in the reactor operation only when its extreme values exceed the threshold value A.sub.max. Only rapidly increasing oscillations cause this extremely critical state to be recognized in good time and to the initiation of suitable countermeasures. Apparently, it is assumed that slowly increasing oscillations inherently decay by themselves and normally do not require a SCRAM. To be specific, that prior art provides as counter-measure only to damp the oscillation by means of rapidly moving in virtually all the control rods (total SCRAM). That is to say, apart from the SCRAM, this strategy provides no further measure for damping the oscillation and does not reduce the probability of the SCRAM either, which constitutes a considerable intervention in the reactor operation. Instead, in the event that there is a rapidly increasing oscillation, damping only takes place earlier (i.e., below A.sub.max). As a result, only the thermal loading of the fuel elements is reduced. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a method and device for operating a reactor in an unstable state, which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which improves the oscillation detection and damping so as to entirely obviate any SCRAM, i.e., to manage without an intervention in the reactor operation, or with an intervention which is the least disturbing. It is a further object to allow the monitoring of the critical state with a system and method which is least susceptible to interference. With the foregoing and other objects in view there is provided, in accordance with the invention, in a reactor operated in accordance with operationally dependent input parameters, a method of operating the reactor which is unstable as a result of an oscillation of an internal physical variable, which comprises: measuring the physical variable during at least two oscillations and calculating at least one measured value for a rate of increase of the oscillation; and PA1 deciding, in dependence on the measured value, whether a stabilization strategy is to be initiated with changed input parameters for damping the instability or a reactor operation is to be continued with unchanged input parameters. PA1 a system selection stage, a plurality of region selection stages connected to the system selection stage, a given number of region monitoring stages connected to each the region selection stage, and a sensor stage connected to each the region monitoring stage with a plurality of sensors strategically disposed in regions of a reactor core of a boiling water reactor, wherein PA1 a) a plurality of sensors disposed in a plurality of regions of a reactor core of a boiling-water reactor, the sensors measuring a physical variable of the reactor core and outputting output signals, the output signals of a plurality of the sensors of a given region being combined into an associated region signal; PA1 b) a plurality of evaluation stages each receiving a respective region signal, the evaluation stages identifying in the region signal an occurrence of extreme values of the physical variable and, given an oscillation of constant frequency, determining a rate of increase of the extreme values in the respective region; and PA1 c) at least one monitoring stage receiving output signals from the evaluation stages, the monitoring stage setting an alarm signal when the extreme values of a predefined number of regions satisfy a local monitoring criterion which depends on the rate of increase of the extreme values. By measuring the physical variable (i.e., the neutron flux, in the case of the thermally-hydraulically induced oscillations) the invention provides for the formation of local measured values in a plurality of regions of the reactor core, the measured values being assigned to the respective regions. The monitoring of the measured values leads to the formation of a current alarm stage from among a hierarchy of alarm stages with associated monitoring criteria, and the selection of the highest alarm stage, whose monitoring criterion is satisfied by the measured values in a predefined minimum number of the regions. (The monitoring criterion can in this case be composed of a plurality of individual conditions, for example the exceeding of separate threshold values for the amplitude and for the rate of increase of the extreme values.) Depending on the current alarm stage, a stabilization strategy is then initiated. As a stabilization strategy which belongs to a low-ranking alarm stage, provision is made to intervene in the operational control and regulation of the reactor only so as to block a removal of the control rods, as is envisaged in the case of an operational increasing of the reactor power: the power of the reactor cannot then be raised by the operating personnel of the reactor; instead only such control commands which correspond to the control of the reactor to a constant or decreasing power become effective in the reactor control system. In at least one higher-ranking alarm stage, provision is made as stabilization strategy for a plurality of control rods to be introduced into the core in the sense of a reduction in the reactor power (alarm stage I). Advantageously, at least two higher-ranking alarm stages (alarm stage II and alarm stage III) are provided, in alarm stage II only a plurality of control rods, corresponding to a fraction of the total number, being moved into the core slowly and in such a way as corresponds to an operational reduction in the power (that is to say the reactor control system performs an operational reduction in the power, even if, for example, a higher power consumption would intrinsically require a higher reactor power and the operating personnel wish to increase the reactor power). In the second higher-ranking alarm stage (alarm stage III)--in a manner similar to the case of a total rapid shutdown of the reactor (total SCRAM)--control rods are moved in rapidly, however likewise not all thereof but only some of the control rods being involved ("partial SCRAM"). A total SCRAM is then no longer necessary, but an option for an alarm stage IV which triggers the SCRAM can be retained. In particular, during the monitoring of the measured values, at least two periods of the oscillation are evaluated, so that the reactor is therefore initially further operated in an unchanged manner, although an oscillation is already indicated. Furthermore, a method of operating a reactor which is unstable as a result of oscillation of a physical variable occurring in the core makes provision, by measuring the physical variable, for forming a measured value which registers the rate of increase of the oscillation (if appropriate, also further measured values). Depending on this measured value, a decision is made as to whether a stabilization strategy should be initiated in order to damp the instability or the reactor is initially further operated in accordance with measured values entered as a function of operation. In particular, in this case the reactor can continue to be operated for at least two more oscillations during the measurement of the rate of increase, without an intervention being made in the reactor control system--provided that no measured value reaches a threshold value which calls for the initiation of a total SCRAM. Thus, for example, it is possible that when a threshold value A.sub.max for the oscillation amplitudes is exceeded, the SCRAM--corresponding to the highest alarm stage IV--is initiated only at high rates of increase, but at low rates of increase the reactor is still operated with relatively high amplitudes, since in the case of amplitudes which are growing so weakly, a SCRAM which is initiated only later (in the event that the oscillation then does not intrinsically decay) still has sufficient time to become effective before A.sub.th is reached. A threshold value, dependent on the rate of increase, is preferably predefined for the extreme values of the oscillating physical variable, and the stabilization strategy is triggered if the extreme values exceed this threshold value. However, a threshold value for the rate of increase can also be predefined, the stabilization strategy then being initiated when the rate of increase exceeds this threshold value. In a similar embodiment of the invention, a number of oscillations can be predefined, the number depending on the rate of increase, and the stabilization strategy can then be triggered only when the oscillation of the physical variable persists over the duration of these oscillation periods. In accordance with an added feature of the invention, a plurality of stabilization strategies are provided, from which the stabilization strategy to be triggered is selected as a function of the rate of increase. The (unstable) state of the reactor core is monitored with a plurality of sensors which are strategically distributed about the core. The sensor locations are divided into a plurality of regions of the reactor core and the sensors measure the behavior of the physical variable in those regions. The output signals of the sensors are combined into a number Mp of region channels and each region channel is assigned a region and sensors arranged therein for generating a region signal. The region signals are then combined into a number P of system channels, with a plurality of region channels being assigned to a system channel, in that they generate a system signal. The system signals are finally assigned to an output channel and they generate an output signal. By means of monitoring stages and selection stages, in this case an alarm output signal is set in the output signal as soon as, at least in a predefined number N.sub.p of the system channels, particularly in a minimum number N.sub.mp of region channels of the system, a monitoring criterion is satisfied over a plurality of oscillation periods. In this case, the output signal of each sensor influences a maximum of one single region signal and each region signal influences a maximum of one system signal. The region signals of a system channel are in each case formed from the output signals of sensors which are located in regions which are distributed over the cross section of the reactor core in such a way that the regions which are adjacent to such a region contain sensors whose output signals are assigned to region channels of other system channels. The invention thus effectively dispenses with multiple evaluations and region overlaps. Although each individual fuel element is no longer as precisely monitored as in the Watford et al. patent, experience and model calculations with unstable states have shown that it is always relatively large parts of the reactor, but not isolated fuel elements, which begin oscillating. In other words, fine resolution of the measured value registration is not necessary. In addition, the redundancy and interference immunity of the registration is increased. With the above and other objects in view, there is further provided, in accordance with the invention, a device for monitoring a reactor core of a boiling-water reactor with regard to local oscillations of a physical variable causing an unstable state of the reactor. The device comprises: a) measured signals supplied by the sensors to a respective the region monitoring stage are combined into a region signal for the physical variable; each the region signal is monitored in the respective the region monitoring stage in accordance with a monitoring criterion, and a region signal containing a region monitoring signal is output by each region monitoring stage; PA2 b) each region signal is connected to at least one of the region selection stages, and the region selection stages forming respective system monitoring signals from a predefined minimum number of region monitoring signals; and PA2 c) the region selections stages each outputting a respective system monitoring signal to the system selection stage, and the system selection stage outputting an output monitoring signal according to a predefined minimum number of systems. There is further provided, in accordance with the invention, a device for monitoring a reactor core of a boiling-water reactor with regard to a state which is unstable as a result of local oscillation of a physical variable in the reactor core, comprising: In other words, there is provided a system selection stage, a number P of region selection stages, for each region selection stage a number Mp of region monitoring stages and for each region monitoring stage a sensor stage having a plurality of sensors which are arranged inside a region of the core and are assigned to this region monitoring stage. The device is constructed in such a way that the sensors which are respectively assigned to a region monitoring stage supply measured signals for the physical variable which are combined into a region signal, and each region signal is monitored in accordance with a monitoring criterion in the region monitoring stage assigned to the sensors. Each region monitoring stage supplies a region signal which contains a region monitoring signal. Each region monitoring signal is connected to at least one region selection stage which forms a system monitoring signal from a predefined minimum number of region monitoring signals. Each system monitoring signal is then fed to the system selection stage; the latter supplies an output monitoring signal by means of a predefined minimum number of system monitoring stages. The system sensors which are strategically distributed about a plurality of regions of the reactor core for measuring the physical variable. The output signals of a plurality of sensors of a region are combined into an associated region signal. Each region signal is assigned an evaluation stage, which identifies in the region signal the occurrence of extreme values of the physical variable (in particular over a plurality of oscillation periods) and, given an oscillation of constant frequency and appropriate duration, determines the rate of increase of the extreme values in this region. The evaluation stages are assigned at least one monitoring stage which sets an alarm signal as soon as the extreme values at least in a predefined number of regions satisfy a local monitoring criterion which depends on the determined rate of increase. With a view to the proposed stabilization criteria, a device for monitoring the local oscillations can contain sensors for measuring the physical variable, which sensors are arranged in a plurality of regions of the reactor core, and the output signals of a plurality of sensors of a region being combined into an associated region signal. Each region signal is then assigned an evaluation stage which identifies the occurrence of an oscillation of constant frequency in the region signal. The evaluation stages are assigned an output monitoring stage which selects an alarm stage from a hierarchy of alarm stages in accordance with predefined monitoring criteria for the oscillations identified in at least a predefined number of region signals. In this case, the output monitoring stage, corresponding to the selected alarm stage, defines a point in time (or at least the criteria for the point in time) at which an emergency instruction is output to initiate a stabilization strategy corresponding to the alarm stage. This point in time can be predefined, for example, by means of a number of oscillation periods which are allowed to elapse before the initiation of a stabilization measure. However, by this means it can also be defined that, depending on the instantaneous current values (for example current values of the rate of increase) a threshold value (for example a threshold value for the amplitude) is defined, which leads to the triggering of the stabilization measure at a later point in time, at which a monitored current value (e.g., the amplitude) then exceeds this predetermined threshold value. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a method and device for operating a reactor in an unstable state, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings.
060977795	description	DETAILED DESCRIPTION FIG. 1 is a schematic, partial cross section, illustration of a boiling water reactor 100 including a reactor pressure vessel (RPV) 102 and a bridge 104. RPV 102 has a generally cylindrical shape and is closed at one end by a bottom head 106 and at its other end by removable top head (not shown). A top guide 108 is spaced above a core plate 110 within RPV 102. A shroud 112 surrounds core plate 110 and is supported by a shroud support structure 114. An annulus 116 is formed between shroud 112 and the wall of RPV 102. A baffle plate 118, which has a ring shape, extends around RPV 102 between shroud support structure 114 and the wall of RPV 102. RPV 102 is supported by an RPV support structure 120 and RPV 102 extends into an upper containment 122. Upper containment 122 and RPV 102 are, of course, filled with water. A water level 124 is shown as being just below bridge 104. RPV 102 is shown in FIG. 1 as being shut down with many components removed. For example, and in operation, many fuel bundles and control rods (not shown) are located in the area between top guide 108 and core plate 110. In addition, and in operation, steam separators and dryers and many other components (not shown) are located in the area above top guide 108. Top guide 108 is a latticed structure including several top guide beams 126 defining top guide openings 128. Core plate 110 includes several recessed surfaces 130 which are substantially aligned with top guide openings 128 to facilitate positioning the fuel bundles between top guide 108 and core plate 110. Fuel bundles are inserted into the area between top guide 108 and core plate 110 by utilizing top guide openings 128 and recessed surfaces 130. Particularly, each fuel bundle is inserted through a top guide opening 128, and is supported by core plate 110 and top guide beams 126. FIG. 2 is a schematic, top view, illustration of a large fuel assembly 132 having four large fuel bundles 134 and a large control rod 136 positioned through one of top guide openings 128. Each large fuel bundle 134 includes a handle 138 adjacent its top end 140, and includes a bundle channel substantially encapsulating a plurality of fuel rods (not shown). Large control rod 136 includes a central portion 142 having four radially extending blades 144. Central portion 142 and blades 144 define four fuel bundle receiving channels 146. Each fuel bundle receiving channel 146 is sized to receive one large fuel bundle 134, and blades 144 facilitate providing poison control between adjacent fuel bundles 134. A control rod drive (not shown) is coupled to control rod 136 for moving control rod 136 relative to top guide 108. For example, during reactor operation, the control rod drive fully inserts control rod 136 within the area between top guide 108 and core plate 110, and a top portion 148 of control rod 136 is substantially adjacent top guide 108. Alternatively, the control rod drive also may withdraw, entirely or partially, control rod 136 from the area between top guide 108 and core plate 110 so that top portion 148 of control rod 136 is spaced from top guide 108. Several channel fasteners 150 are coupled to each fuel bundle 134 and facilitate supporting fuel bundles 134 within receiving channels 146. Particularly, channel fasteners 150 each include a spring, and are positioned so that the each spring pushes one of large fuel bundles 134 against a respective corner 152 of top guide opening 128, i.e., all four fuel bundles 134 are pushed against different respective corners 152 of top guide opening 128. Large fuel bundles 134 have substantially twice the pitch as standard size fuel bundles (not shown) in a conventional BWR fuel configuration. As described above, maximum channel integrated power (i.e., highest radial peaking factor) is greater utilizing large fuel bundles 134 than for a core loaded with conventional size fuel bundles. While this increased pitch is desirable, large fuel bundles 134 substantially prevent performing sub-bundle shuffling, which often is desirable. Moreover, large fuel bundles 134 require larger bundle channels than standard size fuel bundles, and such larger bundle channels are expensive, add more parasitic material in the core region, and may have problems with bend and bow when irradiated. In accordance with one embodiment of the present invention, and to obtain the benefits of large fuel bundles, i.e., increased peak and reduced control rods and control rod drives, without suffering the burdens identified above, substantially standard size fuel bundles are utilized. Particularly, a nuclear reactor core includes large control rods and conventional size fuel assemblies. Each large control rod is sized to provide poison control (e.g., negative reactivity) for sixteen conventional size fuel assemblies, which are configured as four large bundles. The conventional size fuel assemblies are positioned in a "K" lattice configuration to facilitate minimizing the number of control rod drives and control rods. FIG. 3 is a schematic, partial top view, illustration of a reactor pressure vessel 158 including large control rods 160 and standard size fuel bundles 162 configured in accordance with one embodiment of the present invention. Large control rods 160 each include a central portion 164 having four radially extending blades 166. Large control rods 160 are positioned in a staggered configuration, and fuel bundle receiving channels 168 are formed by blades 166 of adjacent large control rods 160. One group, or set, of four standard size fuel bundles 162 is positioned in each fuel bundle receiving channel 168, i.e., four standard size fuel bundles 162 are positioned between each adjacent large control rod 160. This configuration is sometimes referred to herein as an "F-lattice" configuration. Large control rod blades 166 facilitate providing poison control between adjacent groups of fuel bundles 162. FIG. 4 is a schematic, more detailed, top view of sixteen standard size fuel bundles 162 and one large control rod 160. Each fuel bundle 162 includes a handle 170 extending from its top end 172. Fuel bundles 162 are configured in four groups 174A, 174B, 174C, and 174D, and each group 174A, 174B, 174C, and 174D includes a 2.times.2 matrix of bundles 162. Each group 174A, 174B, 174C, and 174D also is positioned within a respective fuel bundle receiving channel 168. With respect to each group 174A, 174B, 174C, and 174D, channel spacers 176 are coupled to each fuel bundle 162 to substantially space each fuel bundle 162 from an adjacent fuel bundle 162. In addition, spring and guard assemblies 178 are coupled to fuel bundles 162 to facilitate supporting fuel bundles 162 within bundle receiving channels 168. For example, and referring only to group 174A of fuel bundles 162, four spring and guard assemblies 178 are substantially centered within group 174A, and are coupled to adjacent corners 180, or central corners, of fuel bundles 162, respectively. Additional spring and guard assemblies 178 are coupled to opposite corners 182, or outer corners, of fuel bundles 162, respectively. It is believed that spring and guard assemblies 178 substantially obviate any need for a top guide to support fuel bundles 162 within reactor pressure vessel 158. Particularly, spring and guard assemblies 178 coupled to central corners 180 of fuel bundles 162 of each group 174A, 174B, 174C, and 174D cooperate with channel spacers 176 to separate and support each fuel bundle 162 within each group 174A, 174B, 174C, and 174D. Spring and guard assemblies 178 coupled to outer corners 182 of group 174A, 174B, 174C, and 174D cooperate to separate and support fuel bundles 162 of adjacent groups 174A, 174B, 174C, and 174D even without a top guide. For example, and referring still to FIG. 4, each group 174A, 174B, 174C, and 174D of fuel bundles 162 is substantially supported within reactor pressure vessel 158 by adjacent groups 174A, 174B, 174C, and 174D of fuel bundles. More particularly, each fuel bundle 162 in group 174A, for example, includes one spring and guard assembly 178 coupled to outer corner 182. Each such assembly 178 is substantially in physical contact with a spring and guard assembly 178 which is coupled to a fuel bundle 162 of an adjacent group of fuel bundles 162. For example, one spring and guard assembly 178 of group 174A is in physical connection with one spring and guard assembly 178 of group 174B. Similarly, one spring and guard assembly 178 of group 174A is in physical connection with one spring and guard assembly 178 of group 174D. In addition, another spring and guard assembly 178 of group 174A is in physical connection with one spring and guard assembly of another adjacent group (not shown) of fuel bundles 162. Accordingly, spring and guard assemblies 178 are coupled to each fuel bundle 162 and separate and support each fuel bundle 162 with respect to an adjacent one of fuel bundles 162. FIG. 5 is a schematic, partial top view, illustration of large control rods 160 and fuel bundles 162 positioned adjacent a large top guide opening 190, or cell opening, which, as described above, is defined by a lattice of top guide beams 192. One of large control rods 160 is substantially centered within top guide opening 190, and sixteen fuel bundles 162, i.e., four groups 174A, 174B, 174C, and 174D of fuel bundles 162 are substantially aligned within top guide opening 190. FIG. 6 is a schematic, more detailed, top view of groups 174A, 174B, 174C, and 174D of standard size fuel bundles 162 and one of large control rods 160 adjacent large top guide opening 190. As described above, a handle 170 extends from top end 172 of each fuel bundle 162. Each group 174A, 174B, 174C, and 174D includes a 2.times.2 matrix of bundles 162 and is positioned within a respective bundle receiving channel 168 of large control rod 160. Channel spacers 176 are coupled to each fuel bundle 162 to space each fuel bundle 162 from an adjacent fuel bundle 162 within each group 174A, 174B, 174C, and 174D, respectively. In addition, spring and guard assemblies 178 are coupled to each fuel bundle 162 and facilitate supporting fuel bundles 162 within bundle receiving channels 168. For example, and referring only to group 174A of fuel bundles 162, four spring and guard assemblies 178 are centered within group 174A, and are coupled to central comers 180 of fuel bundles 162 in group 174A. Additional spring and guard assemblies 178 are coupled to outer comers 182 of fuel bundles 162 in group 174A. In this configuration, large top guide beams 192 provide lateral support for sixteen fuel bundles 162. In addition, spring and guard assemblies 178 and channel spacers 176 cooperate with beams 192 to support and stabilize bundles 162 within respective bundle receiving channels 168. Particularly, and as shown, spring and guard assemblies 178 of outer comers 182 of each group 174A, 174B, 174C, and 174D are in physical connection with either large control rod 160 or top guide beams 192. Accordingly, spring and guard assemblies 178 of outer corners 182 substantially cooperate with control rod 160 and top guide beams 192 to support each group 174A, 174B, 174C, and 174D or bundles 162 within respective bundle receiving channels 168. Spring and guard assemblies 178 are coupled to bundles 162 and positioned above top guide beams 192. In addition, channel spacers 176 also are positioned above top guide beams 192. Such positioning is believed to facilitate positioning groups 174A, 174B, 174C, and 174D of bundles 162 within each bundle receiving channel 168, and facilitates moving individual bundles 162 from a group 174A, 174B, 174C, and 174D. FIG. 7 is a schematic top view illustration of fuel bundles 162 and large control rod 160 positioned adjacent additional fuel bundles 194. Additional fuel bundles 194 are the same as fuel bundles 162 and include a handle 196 extending from a top end 198 thereof. Spring and guard assemblies 178 are coupled to additional fuel bundles 194 and substantially abut spring and guard assemblies 178 coupled to opposite comers 182 of fuel bundles 162 in group 174B and 174C, respectively. In addition, channel spacers 176 are coupled to additional fuel bundles 194 and couple to channel spacers 176 which are coupled to fuel bundles 162 in groups 174B and 174C. FIG. 8 is a cross sectional view along the line A--A shown in FIG. 7. Fuel bundles 194 and 162 each extend above top guide beam 192. One spring and guard assembly 178 is coupled to each fuel bundle 194 and 162, respectively, to support fuel bundles 162 and 194. Each spring and guard assembly 178 includes a channel guard 200 and a spring 202. With respect to fuel bundle 162, channel guard 200 is coupled to top end 172 and extends between fuel bundles 162 and 194. Spring 202 also is coupled to top end 172 of fuel bundle 162, and extends between fuel bundles 162 and 194 so that channel guard 200 is between spring 202 and fuel bundle 162. Similarly, and with respect to fuel bundle 162, channel guard 200 and spring 202 are coupled to top end 198, and extend between fuel bundles 162 and 194 so that springs 202 are substantially adjacent. FIG. 9 is a cross sectional view along the line B--B shown in FIG. 7. Channel spacer 176 includes a male portion 204 and a female portion 206. Male portion 204 is coupled to one fuel bundle 194 and female portion 206 is coupled to one of adjacent fuel bundles 162. Each portion 204 and 206 is positioned above top guide beam 192, and male portion 204 is inserted into female portion 206. FIG. 10 is an exploded schematic top view illustration of two fuel bundles 162 of group 174B adjacent two fuel bundles 194 shown in FIG. 7. Channel spacers 176 each include a male portion 204 and a female portion 206 and are positioned proximate comers 208 of fuel bundles 162 in group 174B and fuel bundles 194. Male portions 204 of each channel spacer 176 interlock with female portions 206 of each channel spacer 176 to separate each fuel bundle 162 from one of adjacent fuel bundles 162 and one of adjacent fuel bundles 192. In addition, male portions 204 and female portions 206 facilitate removing only one of fuel bundles 162 or only one of fuel bundles 194 during refueling. Of course, more than one fuel bundle 162 or more than one fuel bundle 194 may be removed during refueling. FIG. 11 illustrates a large control rod 160 substantially aligned with one of top guide openings 190. Large control rod 160 is substantially free standing, and is positioned within top guide opening 190 so that blades 166 are substantially adjacent top guide beams 192 (only two top guide beams 192 are shown). Accordingly, top guide beams 192 support large control rod 160 within top guide opening 190. In the "K" lattice configuration, however, several large control rods 160 are not substantially aligned with one of top guide openings 190. FIG. 12 illustrates one such large control rod 160 positioned underneath top guide beams 192, but not substantially aligned with a top guide opening 190. As shown, top portions 210 of each blade 166 are configured to be inserted at least partially in top guide beams 192. Particularly, each top guide beam 192 includes a recess 212 sized to receive top portion 210 of one of blades 166. Accordingly, beams 192 substantially support top portions 210 of blades 166, and thus substantially support large control rod 160. Recess 212 also is sized to accommodate any control rod drive overtravel. Particularly, FIGS. 13 and 14 are exploded views of large control rod 160 and top guide beams 192 with large control rod 160 in a lowered position and a raised position, respectively. The above described reactor including large control rods and conventional size assemblies is believed to be operational both with and without a top guide. In addition, such reactor facilitates sub-bundle shuffling, i.e., moving one conventional size fuel bundle at a time, for maximum fuel cycle optimization. Furthermore, such reactor reduces the number of control rod drives by about one-half, as compared to a conventional reactor, and permits refueling four or more conventional fuel assemblies at one time, with an overall reduction in capital cost of the plant and reduced outage time. It is to be understood that the present invention is not limited to practice in reactor 100 and the present invention could be used in may different reactors having many different alternative configurations. Reactor 100 is illustrated by way of example only and not by way of limitation. For example, FIG. 15 is a partial top view illustration of a top guide 290 in accordance with another embodiment of the present invention. Top guide 290 includes a lattice of top guide beams 292 defining large top guide, or cell, openings 294 (only nine cell opening 294 is shown in FIG. 15). Several of top guide beams 292 include chimney partitions 296 extending therefrom, and each cell opening 294 is sized to received sixteen fuel bundles in the "F-lattice" configuration, i.e., four groups 174A, 174B, 174C, and 174D of four fuel bundles 162. Top guide 290 further includes several removable support elements 298 (only one support element 298 is shown in FIG. 15) coupled to top guide beams 292 and extending across cell openings 294. Each support element 298 includes a central portion 300 having four legs 302A, 302B, 302C, and 302D extending therefrom. Each leg 302A, 302B, 302C, and 302D extends from central portion 300 so that an angle between each leg 302A, 302B, 302C, and 302D and an adjacent leg 302A, 302B, 302C, and 302D, is approximately ninety degrees. Each leg 302A, 302B, 302C, and 302D is coupled to one top guide beam 292 so that support element 298 defines four group openings 304A, 304B, 304C, and 304D within each cell opening 294, and each group opening 304A, 304B, 304C, and 304D is sized to receive four fuel bundles, i.e., one group 174A, 174B, 174C, and 174D of fuel bundles 162. FIG. 16 illustrates more clearly a chimney partition 296 extending from an upper surface 306 of one top guide beam 292. FIG. 17 illustrates a support element 298 coupled to top guide beams 292 which do not include chimney partitions 296. Referring now to FIG. 18, support element legs 302A, 302B, 302C, and 302D (only leg 302A is shown in FIG. 17) are coupled to respective top guide beams 292 (only one top guide beam 292 is shown in FIG. 17) utilizing latch assemblies 308 (only one latch assembly 308 is shown in FIG. 17). Referring specifically to support element leg 302A, latch assembly 308 includes a spring-mounted latch member 310 mounted to leg 302A with a spring element 312 so that latch member 310 may move relative to leg 302A. In an extended position (shown in FIG. 18), one end of latch member 310 extends through an engaging element 314 which is coupled to top guide beam 292 and couples leg 302A to top guide beam 292. In a retracted position (not shown in FIG. 18), latch member 310 does not engage engaging element 314 and enables leg 302A to be removed from top guide beam 292. Each additional leg 302B, 302C, and 302D is coupled to additional top guide elements 292 in the same manner. To prepare a reactor for operation, removable support elements 298 are coupled to top guide beams 292 to define several group openings 304A, 304B, 304C, and 304D. One group of four fuel bundles, e.g., fuel bundles 162, is inserted through each group opening 304A, 304B, 304C, and 304D, and the individual fuel bundles are substantially supported within the reactor by top guide beams 292 and support element legs 302A, 302B, 302C, and 302D. Accordingly, removable support elements 298 are believed to obviate any need for channel spacers 176 and spring and guard assemblies 178 between adjacent groups, e.g., groups 174A, 174B, 174C, and 174D, of fuel bundles. In addition, removable support elements 298 do not substantially inhibit the removal of large control rods from the reactor. Specifically, to remove a large control rod, e.g., large control rod 160, aligned with a cell opening 294, the removable support element 298 extending across such cell opening 294 is detached and removed from respective top guide beams 292. The control rod is then lifted through cell opening 294. FIG. 19 is a schematic, partial top view, illustration of a large control rod 320 and sixteen standard size fuel bundles 322 arranged in a F-lattice configuration and positioned adjacent a top guide 324 in accordance with yet another embodiment of the present invention. Top guide 324 includes a lattice of top guide beams 326 defining large top guide, or cell, openings 328 (only one cell opening 328 is shown in FIG. 18). Top guide 324 further includes several top guide beam segments 330A, 330B, 330C, and 330D extending from top guide beams 326 to define group chambers, or group openings, 332A, 332B, 332C, and 332D, within each cell opening 328. Beam segments 330A and 330C are substantially aligned and extend from opposite top guide beams 326. Similarly, beam segments 330B and 330D are substantially aligned and extend from opposite top guide beams 326 so that a line extending between beam segments 330B and 330D is substantially perpendicular to a line extending between beam segments 330A and 330C. Large control rod 320 is substantially aligned with cell opening 328 and, like large control rods 160, includes a central portion 334 having four radially extending blades 336 which define four fuel bundle receiving channels 338. Blades 336 are substantially aligned with respective top guide beam segments 330A, 330B, 330C, and 330D, so that fuel bundle receiving channels 338 are aligned with group openings, 332A, 332B, 332C, and 332D. One group, or set, of four standard size fuel bundles 322 is positioned in each group opening 332A, 332B, 332C, and 332D. Large control rod blades 336 facilitate providing poison control between adjacent groups of fuel bundles 322. Each fuel bundle 322 includes a handle 340 extending from its top end 342. Fuel bundles 322 are configured in four groups 344A, 344B, 344C, and 344D, and each group 344A, 344B, 344C, and 344D includes a 2.times.2 matrix of bundles 322. Each group 344A, 344B, 344C, and 344D also, as described above, is positioned within a respective group opening 332A, 332B, 332C, and 332D. With respect to each group 344A, 344B, 344C, and 344D, channel spacers 346 are coupled to each fuel bundle 322 to substantially space each fuel bundle 322 from an adjacent fuel bundle 322 in such group 344A, 344B, 344C, and 344D. In addition, spring and guard assemblies 348 are coupled to fuel bundles 322 to facilitate supporting each group 344A, 344B, 344C, and 344D of fuel bundles 322 within its respective group opening 332A, 332B, 332C, and 332D. For example, and referring only to group 344A of fuel bundles 322, four spring and guard assemblies 348 are substantially centered within group 344A, and are coupled to adjacent comers 350, or central comers, of fuel bundles 322, respectively. In addition, channel spacers 346 are coupled to each bundle 322 to support each adjacent bundle 322 within group 344A. Spring and guard assemblies 348 and channel spacers 346 cooperate with top guide beams 326 and beam segments 330A and 330D to support group 344A of bundles 322 within respective group opening 332A. Each other group 344B, 344C, and 344D of bundles 322 is similarly supported with respect to group openings 332B, 332C, and 332D. Particularly, spring and guard assemblies 348 and channel spacers 346 cooperate with top guide beams 326 and beam segments 330A and 330B to support group 344B of bundles 322 within respective group opening 332B, spring and guard assemblies 348 and channel spacers 346 cooperate with top guide beams 326 and beam segments 330B and 330C to support group 344C of bundles 322 within respective group opening 332C, and spring and guard assemblies 348 and channel spacers 346 cooperate with top guide beams 326 and beam segments 330C and 330D to support group 344D of bundles 322 within respective group opening 332D. FIG. 20 illustrates top guide beams 326 and beam segments 330A, 330B, 330C, and 330D, without fuel bundles 322 and large control rod 320. FIG. 21 illustrates top guide beams 326 and beam segments 330A, 330B, 330C, and 330D, without fuel bundles 322 and with large control rod blades 336 substantially aligned with respective beam segments 330A, 330B, 330C, and 330D. FIG. 22 illustrates top guide beams 326 and beam segments 330A, 330B, 330C, and 330D, without fuel bundles 322 and with large control rod 320 rotated for removal from the reactor. Particularly, while rotated, large control rod 320 may be lifted through cell opening 328 so that blades 336 extend through respective group openings 332A, 332B, 332C and 332D. The above-described top guide including beam segments 330A, 330B, 330C, and 330D obviates any need for channel spacers 176 and spring and guard assemblies 178 between adjacent groups, e.g., groups 174A, 174B, 174C, and 174D, of fuel bundles, as described with respect to FIGS. 3 through 14. In addition, such beam segments 330A, 330B, 330C, and 330D, do not substantially inhibit the removal of large control rods from the reactor. From the preceding description of the present invention, it is evident that the objects of the invention are attained. Although the invention has been described and illustrated in detail, it is to be clearly understood that the same is intended by way of illustration and example only and is not be taken by way of limitation. Accordingly, the spirit and scope of the invention are to be limited only by the terms of the appended claims.
	abstract	A system for providing radiation protection is provided that includes a garment that contours to an operator's body. The garment protects the operator from radiation. The garment is supported by a suspension component that reduces a portion of weight of the garment for the operator, the garment including a belt, which includes a release mechanism that offers an entry into the garment. In more specific embodiments, the release mechanism is a quick release that allows the operator to disengage from the garment using a single hand movement. The belt can include at least one flexible joint. The belt opens to allow the operator to enter the garment, and the operator, in entering and exiting the garment, is able to limit his contact to components on or near a front of the garment such that the operator can operate the release mechanism for the garment without losing sterility.
	summary
054024552	description	DETAILED DESCRIPTION OF THE INVENTION The present invention provides improved layered storage structure composites for use in the storage of waste materials, especially hazardous, radioactive, and mixed waste materials. The composites can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which are used to house in some manner waste materials. The composite has within its structure a mat that provides improved structural support along with improved radioactive shielding in comparison with other composites that employ concrete mixes with or without metal rebar materials for imparting strength to the composite. The preferred use for the composite is to form a container vessel in which waste materials are placed for storage. The composite material of the present invention contains a fibrous mat layer that is prepared with a concrete-based material within its matrix. The combination of the fibrous mat, preferably made of metal fibers, and the concrete-based material that fills the void spaces within the mat matrix, provides for both a superior strength and shielding composite material for a containment system or container vessel. An embodiment of the invention is depicted in FIGS. 1 and 2 which shows a container vessel 10 in cross-sectional and isometric views. The vessel 10 is used to store waste containers 12 that contain waste materials such as hazardous, radioactive, or mixed waste materials. The vessel 10 is made of the composite layered structure of the present invention. The vessel 10 is outfitted with a mat 20. The mat 20 preferably encompasses the containers 12 along the entire periphery of the area 26 in which the containers are housed. In certain embodiments, a portion of the area 26 can be exposed, or not encompassed by the mat 20, such as the top wall 42 or floor 44 of the vessel 10 as shown in FIG. 2. Preferably, the top wall 42 and floor 44 are made of the same composite layered structure as the rest of the vessel 10. The vessel 10 can have a multitude of geometries. For instance, the vessel 10 can be round, square, or hexagonal among others. Such configurations allow for various packing and storing configurations dependent upon the containment system. The mat 20 is an interwoven matrix of fibrous materials and is shown in more detail in a general cross-sectional view in FIG. 3. The fibers 28 can be made of plastic, ceramic, or metal, such materials that are recycled, or mixtures thereof, and is preferably made of metal fibers such as steel or lead, more preferably stainless steel. The mat 20 is constructed in such a way that the fibers are interwoven to create a tightly woven mesh pad having a thickness of from about 0.6 cm (0.25 in.) to about 10 cm (4 in.), preferably from about 1.2 cm (0.5 in.) to about 7.6 cm (3 in.), more preferably from about 2.5 cm (1 in.) to about 5 cm (2 in.). The individual fibers 28 that constitute the mat 20 are generally from about 10 to about 100 .mu.m, preferably from about 20 to about 60 .mu.m, and more preferably from about 25 to about 40 .mu.m in thickness. The fibers 28 are encased in the concrete 54 matrix. The mat 20 is free-standing in that the interwoven fibers 28 provide support for the mat 20 and the mat 20 can be handled without the fibers 28 becoming disassociated with the mat 20 to a substantial degree. The mat 20 has a fiber volume of from about 1 to about 10, preferably from about 1 to about 5, and more preferably from about 1.5 to about 3, volume percent. An example of such a mat is commercially available from Ribbon Technology Corporation, Gahanna, Ohio. Referring back to FIG. 1, the composite material of which vessel 10 is constructed has at least one layer of support material proximate to the mat 20 to provide both support to the vessel 10 and also for shielding purposes. This layer can be positioned either proximate to the inner face 50 or outer face 52 of the mat 20, preferably proximate to the outer face 52 of the mat 20. Vessel 10 is shown with such an outer layer 22 and with an optional inner layer 18. These layers 18,22 can vary in thickness according to the strength and shielding requirements of the vessel 10, however they are generally from about 2.5 cm (1 in.) to about 15 cm (10 in.), preferably from about 2.5 cm (1 in.) to about 10 cm (4 in.). The layers 18,22 are preferably grout- or concrete-based materials. These materials can include, as dispersed shielding enhancement additives 19, such materials as barite, magnetite, taconite, depleted uranium, and vitrified glass-like materials such as vitrified ash products along with mixtures of these additives. Preferred additives 19 include barite and magnetite. These additives can be admixed with the concrete materials up to about 75, preferably from about 25 to about 75, and more preferably from about 45 to about 70, weight percent. The additives 19 generally are from about 0.5 cm (0.19 in.) to about 1.3 cm (0.5 in.) in particle size, and preferably less than about 5 percent by weight of the additives are below about 100 .mu.m particle size. The vessel 10 is manufactured by positioning the mat 20 into a form and pouring the materials constituting the layer 18 or 22 against the mat 20. The preferred materials for the layer 18 or 22 is a concrete-based mixture containing at least one of the additives 19. It is preferred to limit the amount of water used in the concrete mixture, replacing the water with plasticizer, or superplasticizer, materials. Plasticizers are commonly used materials in the concrete industry and generally extend the slump retention of the concrete mixture, such plasticizers are commercially available from Master Builders, Inc., Cleveland, Ohio as RHEOBUILD 1000 plasticizer. The plasticizers are commonly salts, either calcium or sodium, of beta-naphthalene sulfonate polymers that enable the concrete mixture to meet the ASTM C494 type F concrete specification. It is also preferred to limit the amount of small particle size materials, or "sand-like" particles, in the concrete mixture used for layers 18,22. The amount of particles having a particle size of below about 500 .mu.m, preferably below about 100 .mu.m is below about 10, more preferably below about 5, weight percent of the materials constituting the concrete-based mixture. The concrete-based mixture preferably does not contain sand. These steps are taken to assure that the concrete-based mixture thoroughly permeates the mat 20 matrix, filling the void spaces within the mat 20, and thus producing a sol id cast matrix. Generally, the concrete-based mixture fills at least about 50, preferably at least about 80, and more preferably at least about 95, percent by volume of the void space within the mat 20. The concrete-based mixture generally contains from about 15 to about 40, preferably from about 20 to about 30, weight percent cement; from about 5 to about 15, preferably from about 8 to about 12, weight percent water; from about 10 to about 15 percent by weight fly ash; and at least about 0.5, preferably from about 0.5 to about 0.1, and more preferably from about 0.52 to about 0.8, percent by weight of plasticizer. The concrete-based mixture preferably also contains metallic fibers dispersed within the mixture. These fibers are provided in loose, individual form and can be made from such materials, for example, as steel, including stainless and carbon-coated, along with lead and other metallic materials and their oxides, carbon, and graphite and can further be made of recycled materials of any kind. The fibers are generally about 15 mm (0.62 in.) to about 5 cm (2 in.) in length, about 1-2 mm (0.04-0.08 in.) in width, and about 30 .mu.m in thickness. These metallic fibers are provided in an amount of from about 0.5 to about 3 percent by weight of the concrete-based mixture. These fibers are commercially available from Ribbon Technology Corporation. The incorporation of the fibers provides for an increase strength composite material. The fibers incorporate themselves into the mat 20 matrix during the process of pouring the concrete-based materials. The concrete-based mixture can also contain other materials such as zeolites, activated carbon, sodium silicate, or silica fume, or mixtures thereof. These materials improve the strength and shielding of the composite. The concrete-based mixture is positioned into the mat 20 matrix by the use of vibrators. The larger particle size additives 19 generally cannot enter into the mat 20 matrix, however the concrete mix is made with such a fluidity characteristic that the other concrete-based mixture components are carried into the matrix. The vessel 10 can optionally be manufactured with a inner wall 23 and outer wall 24 that are coated with an impermeable material layer 16. Typical impermeable materials include glass coatings, epoxy coatings, and inorganic coatings such as those containing silica and zirconia. This coating is from about 0.3 cm (0.1 in.) to about 0.6 cm (0.25 in.) in thickness. A further optional layer of the vessel 10 can be a liner 14. The liner 14 is located adjacent to the inner wall 23 and can be made from such materials as steel, lead, and depleted uranium. The various layers that constitute the vessel 10 can also be used for storing purposes in various shapes besides those employed as a vessel. For instance, the layer construction of the present invention can be used to encase several high integrity containers placed in a series or row formation. For instance, in FIG. 4 a cross-sectional view is shown depicting the composite layer structure of the present invention used as a shielding containment system 60. The side walls 40 and top wall 42 are set into place by means of the lugs 32 and held together by means of a bolt 30. In this way, several containers 12 can be set along side one another and the layered shielding walls extended to ensure proper storage. The walls 40,42 of the containment system 60 are typically formed as individual units and must be connected to form an entire containment system 60. As shown in FIG. 5, the walls, such as side wall 40, are interconnected by the use of a joint 62 that is preferably a labyrinth or off-set joint to reduce the streaming of hazardous or radioactive fumes. The improved composite layered structure of the present invention described above can thus be described as having an optional first layer that is the liner 14. Positioned proximate to the liner 14 is an optional impermeable coating 16. Adjacent to the coating 16 is a first concrete-based layer, or inner layer 18, which is located proximate to the mat 20. On the other side of the mat 20 is a second concrete-based layer 22 upon which an optional impermeable layer 16 can be placed. This multilayered composite material displays improved strength and shielding capacity than conventional composite materials which do not contain such a fibrous mat.
	claims	1. A pressure-tube nuclear reactor comprising:an outer shell having at least one shell side wall and a shell tubesheet that cooperate to define an interior to contain a heavy water moderator at a first pressure;a coolant plenum having a plenum cover, at least one plenum side wall and a plenum tubesheet that cooperate to define a plenum chamber to receive a coolant fluid at a second pressure, the second pressure being greater than the first pressure, wherein the plenum tubesheet seals an open end of the outer shell and is in physical contact with the heavy water moderator;a plurality of pressure tubes received within and extending through the interior of the outer shell from the plenum tubesheet to at least the shell tubesheet, each pressure tube configured to releasably retain at least one fuel bundle and having an outer surface in direct physical contact with the heavy water moderator, a first end of each pressure tube being coupled to the plenum tubesheet in fluid communication with the plenum chamber and a second end of each pressure tube fluidly connected to a coolant conduit to enable the coolant fluid to flow between the coolant plenum and each pressure tube and to flow from the nuclear reactor for further processing; andan insulator liner disposed within each pressure tube to inhibit heat transfer between the coolant fluid and the pressure tube. 2. The nuclear reactor of claim 1, wherein the plurality of pressure tubes extend substantially vertically through the interior of the outer shell. 3. The nuclear reactor of claim 1, wherein each insulator liner is porous and a portion of the coolant fluid is retained within each insulator liner. 4. The nuclear reactor of claim 1, wherein each insulator liner comprises at least one ceramic insulator liner loosely received within each pressure tube. 5. The nuclear reactor of claim 1, wherein the plenum is an inlet plenum fluidly connected to at least one coolant supply conduit to receive the coolant fluid and direct the coolant fluid into the plurality of pressure tubes. 6. The nuclear reactor of claim 5, wherein the first end of each pressure tube defines a pressure tube inlet and the second end of each pressure tube defines a pressure tube outlet, each pressure tube outlet fluidly connected to a coolant outlet conduit. 7. The nuclear reactor of claim 6, wherein the coolant outlet conduit comprises at least one riser, a first end of the at least one riser being fluidly connected to a coolant collection header and a second end of the at least one riser being coupled to the plurality of pressure tube outlets, the at least one riser supported by coupling the first end of the at least one riser to a riser support to allow the second end of the at least one riser to accommodate thermal expansion. 8. The nuclear reactor of claim 6, wherein each pressure tube outlet is fluidly connected to the coolant outlet conduit using expansion joints. 9. The nuclear reactor of claim 1, wherein the moderator is pressurized at a first pressure that is less than 1 MPa and the plenum chamber is configured to receive the coolant fluid at a second pressure that is between 8-15 MPa so that the coolant fluid is a subcritical fluid. 10. The nuclear reactor of claim 1, wherein the moderator is pressurized at a first pressure that is less than 1 MPa and the plenum chamber is configured to receive the coolant fluid at a second pressure that is between 23-28 MPa so that the coolant fluid exiting the second end of each pressure tube is a supercritical fluid. 11. The nuclear reactor of claim 1, wherein the plenum chamber is sized to hold a pre-determined volume of coolant fluid so that in use, substantially all radiation shielding for a portion of the nuclear reactor covered by the coolant plenum is provided by the coolant plenum and the pre-determined volume of coolant fluid. 12. The nuclear reactor of claim 1, further comprising an expansion bellows disposed between the coolant plenum and the outer shell to accommodate thermal expansion of at least one of the coolant plenum, the pressure tubes and the outer shell. 13. The nuclear reactor of claim 1, wherein the plurality of pressure tubes have a neutron absorption cross-section between 150-300 mb. 14. The nuclear reactor of claim 1, wherein the second end of each pressure tube is coupled to the outer shell by a respective tube expansion bellows to accommodate for longitudinal growth of each pressure tube. 15. The nuclear reactor of claim 1, wherein the coolant fluid is heavy water or light water. 16. The nuclear reactor of claim 1, further comprising a second coolant plenum comprising a second plenum tubesheet coupled to the second ends of the plurality of pressure tubes and a second plenum chamber to receive the coolant fluid from the plurality of pressure tubes and direct the coolant fluid to the coolant outlet conduit. 17. The nuclear reactor of claim 1, wherein at least one of the insulator liners is formed from ceramic zirconia. 18. The nuclear reactor of claim 1, wherein the insulator liners have a failure pressure at which the insulator liners will fail, and the second pressure is greater than the failure pressure. 19. The nuclear reactor of claim 1, wherein the plenum tubesheet comprises a plenum surface in contact with the coolant fluid in the plenum chamber and an opposed moderator surface in physical contact with the heavy water moderator. 20. The nuclear reactor of claim 1, wherein the plenum tubesheet is convexly curved toward the outer shell. 21. The nuclear reactor of claim 1, further comprising at least one flow regulating element disposed within the plenum chamber to distribute the coolant fluid amongst the pressure tubes. 22. The nuclear reactor of claim 1, wherein a total coolant flow rate divided by the number of pressure tubes in the reactor defines a mean flow rate and wherein a flow rate of coolant fluid through each pressure tube is within 25% of the mean flow rate. 23. The nuclear reactor of claim 1, wherein the insulator liners are removably disposed within the pressure tubes. 24. The nuclear reactor of claim 1, wherein the plenum tubesheet has a wall thickness of between about 40 cm and about 50 cm. 25. A coolant containment system for a nuclear reactor having an outer shell containing a liquid moderator, the coolant containment system comprising:a plenum connectable to an outer shell of a nuclear reactor and having a fluid connection for connecting to a coolant processing system, the plenum comprising a plenum tubesheet and a plenum sidewall extending from the plenum tubesheet to define a plenum chamber and when the plenum is connected to the outer shell the plenum tubesheet seals an open end of the outer shell and is in physical contact with a liquid moderator contained within the outer shell;a plurality of pressure tubes connected at first ends thereof to the plenum tubesheet, the pressure tubes being adapted to receive nuclear fuel bundles and to be mounted within the outer shell and in physical contact with the liquid moderator, and second ends of the pressure tubes fluidly connected to the coolant processing system, the plenum chamber being openable to provide simultaneous access to an interior of the plenum chamber and the plurality of pressure tubes;whereby coolant can be circulated through the coolant processing system, the plenum and the pressure tubes and wherein the moderator is pressurized at a first pressure and the plenum chamber is configured to receive the coolant fluid at a second pressure that is at least 7 MPa greater than the first pressure and the coolant fluid exiting the second end of each pressure tube is a supercritical fluid. 26. The coolant containment system of claim 25, wherein the plenum tubesheet is a pressure barrier between the coolant fluid in the plenum chamber and the moderator and can resist pressure differentials of at least 22 MPa. 27. The coolant containment system of claim 25, further comprising an insulator liner disposed within each pressure tube to inhibit heat transfer between the coolant fluid and the pressure tube. 28. The coolant containment system of claim 27, wherein at least one of the insulator liners is formed from ceramic zirconia. 29. The coolant containment system of claim 25, wherein the plenum chamber is configured to receive the coolant fluid at a pressure that is between 23-28 MPa. 30. The coolant containment system of claim 25, wherein the plenum tubesheet has a wall thickness of between about 40 cm and about 50 cm. 31. A pressure-tube nuclear reactor comprising:an outer shell having, at least one shell side wall and a shell tubesheet that cooperate to define an interior to contain a heavy water moderator at a first pressure;a coolant plenum having a plenum cover, at least one plenum side wall and a plenum tubesheet that cooperate to define a plenum chamber to receive a coolant fluid at a second pressure, the second pressure being greater than the first pressure and the plenum tubesheet is a pressure barrier between the coolant fluid in the plenum chamber and the heavy water moderator contained in the outer shell and can resist pressure differentials of at least 22 MPa; anda plurality of pressure tubes received within and extending through the interior of the outer shell from the plenum tubesheet to at least the shell tubesheet, each pressure tube configured to releasably retain at least one fuel bundle and having an outer surface in direct physical contact with the heavy water moderator, a first end of each pressure tube being coupled to the plenum tubesheet in fluid communication with the plenum chamber and a second end of each pressure tube fluidly connected to a coolant conduit to enable the coolant fluid to flow between the coolant plenum and each pressure tube and to flow from the nuclear reactor for further processing. 32. The nuclear reactor of claim 31, wherein the first end of each pressure tube defines a pressure tube inlet and the second end of each pressure tube defines a pressure tube outlet, each pressure tube outlet fluidly connected to a coolant outlet conduit. 33. The nuclear reactor of claim 31, further comprising an insulator liner disposed within each pressure tube to inhibit heat transfer between the coolant fluid and the pressure tube. 34. The nuclear reactor of claim 33, wherein the insulator liners are removably disposed within the pressure tubes. 35. The nuclear reactor of claim 31, wherein the plenum tubesheet comprises a plenum surface in physical contact with the coolant fluid in the plenum chamber and an opposed moderator surface in physical contact with the heavy water moderator. 36. The nuclear reactor of claim 31, wherein the plenum tubesheet is convexly curved toward the outer shell. 37. The nuclear reactor of claim 31, wherein each insulator liner comprises at least one ceramic insulator liner loosely received within each pressure tube.
	abstract	A method for high spatial resolution imaging of a plurality of sources of x-ray and gamma-ray radiation is provided. High quality mechanically bent diffracting crystals of 0.1 mm radial width are used for focusing the radiation and directing the radiation to an array of detectors which is used for analyzing their addition to collect data as to the location of the source of radiation. A computer is used for converting the data to an image. The invention also provides for the use of a multi-component high resolution detector array and for narrow source and detector apertures.
	abstract	The invention relates to a fuel assembly for a boiling water reactor which is adapted, during operation of the reactor, to allow water to flow upwards through the fuel assembly while absorbing heat from a plurality of fuel rods, whereby part of the water is transformed into steam. The fuel assembly comprises a steam channel through which the steam flows through the fuel assembly. The steam channel (16a, 16b, 16c, 16d) consists of an empty volume which at least extends through part of the fuel assembly. The fuel assembly is designed such that the water and the steam are brought to rotate around the steam channel whereby the water is thrown away from the steam channel whereas the steam which is separated from the water flows upwards through the steam channel.
	claims	1. A nozzle apparatus of a jet pump comprising:a first tubular member;a second tubular member disposed in the first tubular member, apart from the first tubular member;a fluid passage forming member disposed in the first tubular member, and installed to an upper end portion of the second tubular member;a plurality of passage members having one end portion fixed to the first tubular member and having another end portion fixed to the second tubular member, and disposed in the circumferential direction of the nozzle apparatus; andan annular ejection outlet is formed between a lower portion of the first tubular member and a lower portion of the second tubular member;wherein a suction passage formed in each of the passage member, for introducing the suction fluid from outside of the nozzle apparatus to inside of the nozzle apparatus, communicates with an inner region formed in the second tubular member;wherein an opening of the suction passage is formed at the one end portion of each passage member which is fixed to the first tubular member and the opening is opened to an outside region of the nozzle apparatus;wherein an annular driving fluid passage for introducing the driving fluid, across which each of the passage members is disposed, is formed between the first tubular member, the second tubular member and the fluid passage forming member, and communicated with an annular ejection outlet; andwherein an ejection outlet-side portion of the driving fluid passage slopes inward toward the lower end of the nozzle apparatus. 2. The nozzle apparatus of a jet pump according to claim 1, wherein the passage member slopes toward a lower end of the nozzle apparatus as approaching the inner region. 3. The nozzle apparatus of a jet pump according to claim 1, wherein a cross section of the passage member, perpendicular to the axis of the passage member is an oval shape. 4. The nozzle apparatus of a jet pump according to claim 3, wherein the passage member is disposed in a way that the major axis of the oval shape follows the axial direction of the nozzle apparatus. 5. The nozzle apparatus of a jet pump according to claim 1, wherein a surface of the fluid passage forming member, facing the inner region is a curved surface curved from an outlet of the suction passage to a lower end of the fluid passage forming member. 6. The nozzle apparatus of a jet pump according to claim 5, wherein a cross-sectional area of a portion where the curved surface of the fluid passage forming member is formed decreases toward the lower end of the fluid passage forming member. 7. The nozzle apparatus of a jet pump according to claim 1, wherein a cone member in which a cross-sectional area decreases upward is disposed on an upper end of the fluid passage forming member.
047078464	description	DETAILED DESCRIPTION OF THE INVENTION The following detailed description is of the best presently contemplated modes of carrying out the present invention. This description is not intended in a limiting sense but is made solely for the purpose of illustrating the general principles of the invention. Referring now to the drawings in detail, wherein like numerals indicate like elements, there is shown in FIG. 1 the shielding or masking means of the present invention designated generally as 10. The shielding means 10 is substantially square in configuration but may be rectangular or otherwise shaped so as to provide for the blocking of all radiation unnecessary in the creating of an image on an X-ray film of the area under observation or inspection. The shielding means 10 is preferred to be substantially square in shape and approximately 5 inches along a side so as to cooperatively mount immediately adjacent to and in front of the collimator of existing X-ray machinery. The shielding means 10 is comprised of a metallic plate 12 consisting of lead or other suitable material of sufficient thickness to block X-ray radiation. The metallic plate 12 is laminated between two similarly shaped plates of translucent plastic 14, 16 or other suitable material which will not suffer degradation when subjected to intense and repeated exposures to radiation. The thickness of the translucent plastic plates 14, 16 is dependent upon the radiation degradation factor. It is recommended, however, that the plates 14, 16 be at least 0.125 inches thick to withstand normal use in this environment. The plates 12, 14 and 16 may be secured one to the other by means of threadable screws 18 or other suitable means. It is preferred that the threadable screws 18 are inserted through plate 14, then through plate 12 and threaded into previously threaded holes in plate 16 to securedly fasten the plates one to the other. The plates, 12, 14 and 16 are provided with an elongated opening or slot 20 vertically positioned medially along the horizontal measurement of the shielding means 10. A broader or widened portion of the slot 20 is located at the lower or distal end of the slot. This widened portion is approximately twice the width of the upper, narrower portion of the slot 20 and comprises the bottom fourth or one-quarter portion of the slot. The slot 20 provides for the free passage of X-ray radiation from an X-ray radiation generating source to the particular subject area. The remaining portion of the shielding means 10 effectively blocks all other radiation emanating from the X-ray radiation generating source which is not required to form an image on an X-ray film of the area under observation or inspection so as to prevent the subject from exposure to unnecessary and excessive radiation. Referring to FIG. 4, the shielding means 10 is provided with a mounting means 22 adapted to be placed and maintained in front of an X-ray radiation discharge opening of an X-ray machine, for example, a collimator (shown in dotted lines). The mounting means 22 may be permanently secured to the X-ray machine through the holes 23 provided in the top and bottom flanges of the mounting bracket 24 which supports the mounting means. The mounting means 22 is attached to the support bracket 24 by an upper and a lower horizontally placed track which securely hold the cooperating surfaces of and the mounting means in a fixed position and prevent it from moving laterally or outwardly away from the collimator. The lateral positioning is maintained by an abutment wall along the distal side of the bracket 24 and a spring and release tab (34) at the track entrance, both elements being well known in the mechanical arts and similar to the abutment wall and spring and release tab described later in connection with the mounting of the shielding means 10. The shielding means 10 is placed in the mounting means 22 by slidably moving it into the tracks 26, 28 and up against the abutment wall 30. This placement is facilitated by a guide means placed at an angle to the tracks 26, 28 so that the shielding means 10 more easily aligns with the tracks. The shielding means 10 is only loosely held in place by the tracks 26, 28 and rests against a retaining means 33 vertically oriented along the wall 30 to keep the shielding means 10 substantially aligned with the mounting means 22. To retain the lateral positioning of the shielding means 10 a spring member (not shown) is used which operates in conjunction with the spring and release tab 34. The mounting means 22 also includes a second set of tracks 36, 38 which are positioned at an angle of 90 degrees to the first set of tracks 26, 28 so that a transparent cover plate 40 may slide into the tracks 36, 38. The cover plate 40 is fixedly secured in the tracks 36, 38 by a threaded fastener 42. Looking now at FIG. 2, the cover plate 40, also of a preferred substantially rectangular or square configuration, is shown secured in place by the fastener 42 between the tracks 36, 38. The cover plate 40 may be constructed of plastic or other appropriate material suitable for the use and environment. In its preferred embodiment the transparent cover plate 40 is provided with a rectangular slot medially disposed along its horizontal measurement to avoid interference with the X-ray radiation emissions passing through the elongated opening 20 of the shielding means 10. The cover plate 40 will preferably also be provided with one or two annular openings or finger holes 44, 46 to facilitate manual placement and removal of the plate. The cover plate 40 also functions to retain the shielding means 10 in substantial juxtaposed alignment with the collimator as the X-ray machine may be moved or tilted for specific positioning. Alternatively, the shielding means 10 may be placed inside the mounting means 22 within the opening formed by the tracks 26, 28 and the wall 30. The cover plate 40 is then inserted to hold the shielding means 10 in position in front of the collimator. FIG. 2 demonstrates the operation of the full spine shielding means of the present invention. The shielding means 10 is placed in the mounting means 22 immediately adjacent and in front of the X-ray machine collimator, shown in the center background of the drawing as the bisecting openings in the X-ray screen. As can be readily observed from the drawing, the shielding means 10 blocks all X-ray radiation emitted from the collimator field except for the radiation which freely passes through the elongated slot 20 of the shielding means. The image created by the passed X-ray radiation is shown on the X-ray film 48 displaying the image 50 of the elongated slot 20 of the shielding means 10. In its desired application, such shielding or masking will result in exposure of the entire spine but limited to a narrow area along either side of the cervical, dorsal and lumbar regions. The narrow strip widens at the bottom, as does the elongated slot 20, to encompass the entire pelvic region. Such shielding or masking significantly reduces excessive exposure of the human body, or the particular subject area of the body to be observed or inspected, to X-ray radiation. FIG. 3 shows the shielding means 10 in conjunction with additional shielding and filtering means. The transparent plate 52 including a wedge-type X-ray filtering means 54 and reproductive gland shielding means 56, 58 can be used with the shielding means 10 of the present invention to provide greater protection from excessive X-ray radiation primarily in the pelvic region. The plate 52, similarly to cover plate 40, is positioned and secured using the tracks 36, 38 and the fastener 42 as described above. The plate 52 also functions to retain the shielding means 10 in substantial juxtaposed alignment with the collimator of the X-ray machine as it is fixedly secured in tracks 36, 38 by threaded fastener 42 in similar fashion and arrangement as is cover plate 40. Reference may be had to U.S. Pat. Nos. 4,266,139 and 4,472,637 for detailed descriptions of the wedge-type filtering means 54 and the gland shielding means 56, 58 referred to herein other than the present invention producing structural relationships in accordance with the invention. The gland shielding means, when positioned in front of the widened portion of the elongated slot 20 of the shielding means 10 will reduce the effective radiation to the pelvic region in the area of the reproductive glands so as not to expose that particular area of the body to excessive radiation. Thus the present invention is capable of effectively reducing the exposure of particular areas of the human body to excessive unnecessary radiation from an X-ray radiation generating source. This is accomplished by confining the necessary radiation to a particular area and blocking all other radiation from contact with the subject. The radiation necessary to create an image on an X-ray film will pass through the opening in the shielding or masking means while all other radiation is substantially blocked by the shield. The resultant X-ray will be exposed only in the areas coextensive with the elongated slot of the shielding or masking means. Additionally, if a gland shield is used, the area coextensive with that shield will also remain unexposed as the glandular area has been effectively protected from the X-ray radiation by blocking such radiation with the shield. A significant step in reducing unnecessary exposure to X-ray radiation will be accomplished with use of the present invention. The present invention may be embodied in other specific forms without departing from the spirit or essential attributes thereof and, accordingly, reference should be made to the claims rather than to the specification as indicating the scope of the invention.
	claims	1. A method for collecting and designating data regarding an operating condition of a portion of a nuclear reactor core, the method comprising:positioning a first linear array of gamma thermometer (GT) sensors in a first instrument housing, wherein the GT sensors are arranged asymmetrically along the first linear array;positioning a second linear array of GT sensors in a second instrument housing, wherein the GT sensors are arranged asymmetrically along the second linear array and wherein the second linear array of GT sensors is asymmetrical with respect to the first linear array of GT sensors;positioning the first instrument housing in the reactor core at a first core location and positioning the second instrument housing at a second core location symmetrical with respect to the first core location, wherein a line of symmetry of the core extends through a center of the core and the first core location, at which is positioned the first instrument housing and first linear array of GT sensors, and the second core location, at which is positioned the second instrument housing including a second linear array of GT sensors, and the second core location is a same horizontal distance from the line of symmetry as is the first instrument housing, wherein the horizontal distance is along a line perpendicular to the line of symmetry;collecting core condition data from at least one of the GT sensors in the first linear array of GT sensors,designating the collected core condition data as equivalent to core condition data at an elevation on the second linear array that is the same as the elevation of the at least one of the GT sensors in the first linear array; andapplying the designated core condition data and collected core condition data as overall core condition data. 2. The method as in claim 1 wherein at least one of the GT sensors in the first linear array is at a first elevation and the second linear array does not have a GT sensor at the first elevation. 3. The method as in claim 2 wherein the elevation includes a detector for a Local Power Range Monitor (LPRM) adjacent the second linear array, and the method includes calibrating the detector using the collected data from the at least one GT sensor on the first linear array at the same elevation. 4. The method as in claim 1 wherein the second linear array includes a majority of the GT sensors in a lower half of the array and the first linear array includes a majority of the GT sensors in an upper half of the array, and at least one of said GT sensors in the lower half of the array in the first linear array is adjacent each of a plurality of detectors of a first LPRM adjacent the first linear array and at least one of the plurality of detectors in an upper half of the first LPRM has no adjacent GT sensor on the first linear array at the same elevation, and at least one of said GT sensors in the upper half of the array in the second linear array is adjacent each of a plurality of detectors of a second LPRM adjacent the second linear array and at least one of the plurality of detectors in a lower half of the second LPRM has no adjacent GT sensor from the second linear array at the same elevation. 5. The method as in claim 1 wherein the first core location and second core location are at a common distance from a line of symmetry extending through the core. 6. A method for collecting and designating data regarding an operating condition of a portion of a nuclear reactor core, the method comprising:positioning a gamma thermometer (GT) sensor at a first position in the reactor core;collecting core condition data from the GT sensor at the first position,designating the collected core condition data as equivalent to core condition data at a second position in the nuclear core, wherein the second position is at substantially the same elevation in the core as the first position, and the first position and the second position are each a same distance from a line of symmetry which extends horizontally through a center of the core, wherein the horizontal distance is along a line perpendicular to the line of symmetry; andapplying the designated core condition data and collected core condition data as overall core condition data. 7. The method as in claim 6 wherein the second position includes a detector for a Local Power Range Monitor (LPRM) and the method includes calibrating the detector for the LPRM using the collected core condition data. 8. The method as in claim 6 wherein the first position and second position are on opposite sides of the line of symmetry. 9. A method for collecting and designating data regarding gamma flux at positions in a nuclear reactor core, the method comprising:positioning a gamma thermometer (GT) sensor at a first position in the reactor core;collecting gamma flux data from the GT sensor at the first position, wherein the gamma flux data is representative of a gamma flux level at the first position in the core;determining a second position in the core at the same elevation in the core as the first position and at which is equidistant along a line perpendicular to a line of symmetry extending horizontally and through a center of the core,designating the gamma flux data as data collected at the first position as being representative a gamma flux level at the second position in the core; andapplying the designated core condition data and collected core condition data as overall core condition data. 10. The method of claim 9 wherein the core lacks a gamma thermometer (GT) sensor at the second position.
044477341	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a radiation-shielding transparent material and a method of producing the same, which material shields radioactive rays such as neutron beams and .gamma.-rays leaking from nuclear reactors, cyclotrons, or the like. 2. Description of the Prior Art Neutron beams and .gamma.-rays leaking from nuclear reactors, cyclotrons, or the like collide with surrounding substances and cause radiation which may be hazardous to people and apparatuses exposed thereto. The leaking neutrons are in any form of high-speed neutrons, low-speed neutrons, and thermal neutrons. To moderate the high-speed neutrons, use of elements with small atomic numbers and compounds thereof has been known, such as hydrogen, helium, lithium, beryllium, boron, carbon, nitrogen, water, heavy water, and the like. Effective moderation of the high-speed neutrons is achieved by their collision with hydrogen (with an atomic number 1) having a small mass which is similar to that of neutrons, so that materials having a high concentration of hydrogen are very effective in moderating the high-speed neutrons. Water whose molecule has two hydrogen atoms and one oxygen atom is a least expensive yet very effective moderator for shielding the high-speed neutrons. More particularly, the high-speed neutrons are moderated by collision with water and converted into low-speed neutrons and thermal neutrons. However, the elements with a small atomic number and a small mass and compounds thereof are not effective in shielding .gamma.-rays, and only elements with a large atomic number and compounds thereof are effective in shielding .gamma.-rays, such as tungsten, lead, thallium, bismuth, tantalum, thorium, plutonium, and the like. The low-speed neutrons and thermal neutrons are moderated by elements having a large cross section for neutron absorption, such as boron, cadmium, indium, and the like, so that the low-speed neutrons and thermal neutrons are converted by the moderation into .gamma.-rays having an energy of about 0.42 MeV, whereby the overall energy of the leaking radiation is attenuated. Conventionally, heavy concrete containing a moderator, such as iron, lead, barium, metal hydride, serpentine, boron, and the like has been used to shield nueclear reactors, cyclotrons, and the like. The heavy concrete is highly effective in absorbing .gamma.-rays but not so effective in moderating neutrons which are leaking from the nuclear reactors, cyclotrons, or the like. However, no materials capable of effectively moderating both .gamma.-rays and neutrons without being damaged thereby have been found yet. Lead glass has been used as a material for checking windows of nuclear reactors, cyclotrons, and the like, and the lead glass has been known as an effective radiation-shielding transparent material. Nevertheless, the transparent lead glass as the radition-shielding material has shortcomings in that the lead glass is very costly, i.e., one hundred million yen per several cubic meters thereof; that the lead glass is brittle when being machined, so that it has been difficult to machine a lead glass member into desired dimensions with high accuracy; and that the lead glass is coloured with the increase of lead content therein and the transparency thereof is reduced by the colouring. Thus, there have been no radiation-shielding transparent materials, except the lead glass, which are suitable for total absorption calorimeters for measuring the total energy of .gamma.-rays and for shielding nuclear reactors and cyclotrons. The lack of radiation-shielding transparent materials overcoming the aforesaid shortcomings of the lead glass has seriously hampered the research and development of the aforesaid apparatuses of nuclear industries. As regards radiation-shielding materials which are less costly than the lead glass, a solution of zinc bromide (ZnBr.sub.2) has been known, but such a solution has shortcomings in that long-term chemical stability thereof is low and that the transparency thereof is gradually deteriorated. Accordingly, such solutions are seldom used now due to the aforesaid shortcomings. Thus, there has been a pressing need for development of a radiation-shielding transparent material overcoming the shortcomings of the lead glass, so as to further expand the practical applications of radiation-related apparatus: for instance, shielding of nuclear reactors, instruments for measuring radiations such as .gamma.-rays and neutrons, and medical apparatuses using x-rays and .gamma.-rays. SUMMARY OF THE INVENTION The inventors noted the aforesaid points and carried out various studies on radiation-shielding transparent materials suitable for total absorption calorimeters, which materials are stable when being exposed to irradiation of any of .gamma.-rays, x-rays, electron beams, and neutron beams without being damaged thereby. As a result, the inventors have found that a radiation-shielding transparent material can be produced by using a transparent heavy liquid prepared by deoxidizing an aqueous solution of organic thallium compound such as thallium formate and thallium malonate, which material can be used in any of the aforesaid apparatuses. An object of the present invention is to provide a radiation shielding transparent material comprised of a heavy liquid prepared by deoxidizing either an aqueous solution of thallium formate or an aqueous solution of thallium formate and thallium malonate, which material has a density of 2.5 to 4.3 g/cm.sup.3, a radiation length of 3.8 to 1.9 cm, a transmission of not less than 93%, preferably 95 to 99.5%, for light of 400 nm wavelength. Another object of the present invention is to provide a method of producing a radiation-shielding transparent material, comprising steps of separately deoxidizing thallium formate and distilled water, and mixing the thus deoxidized thallium formate and deoxidized distilled water in a non-oxidizing atmosphere at a rate of 300 to 670 grams of thallium formate per 100 cubic centimeters of distilled water, so as to produce a heavy liquid having a density of 2.5 to 3.3 g/cm.sup.3, a radiation length of 3.8 to 2.5 cm, and a transmission of not less than 93%, preferably 95 to 99.5%, for light of 400 nm wavelength. A further object of the present invention is to provide a method of producing a radiation-shielding transparent material composed of heavy liquid, comprising steps of deoxidizing thallium formate, thallium malonate, and distilled water separately; and mixing the thus deoxidized thallium formate, thallium malonate, and distilled water at a rate of 300 to 800 grams of thallium formate and thallium malonate per 100 cubic centimeters of distilled water, so as to produce a heavy liquid having a density of 2.5 to 4.3 g/cm.sup.3 (preferably 3.2 to 4.3 g/cm.sup.3), a radiation length of 3.8 to 1.9 cm (preferably 3.3 to 2.5 cm), and a transmission of not less than 93% for light of 400 nm wavelength. The heavy liquid thus produced can be used as a radiation-shielding transparent material in lieu of the conventional lead glass. The radiation-shielding transparent liquid material of the present invention thus produced is featured in that the optical properties and radiation properties thereof are equivalent to or superior to those of the lead glass; that absorption of .gamma.-rays and neutrons are both high; that the resistivity thereof against radiation damage is considerably higher than that of the lead glass; that cost thereof is noticeably lower than that of the lead glass; and that the freedom of shape and dimension thereof is high.
050200836	abstract	An X-ray mask for manufacturing chips is produced by forming an X-ray transparent semiconductor membrane with gaps and including X-ray transparent material in the gaps. In one embodiment the opaque material is formed by sputtering Pt onto the semiconductor material to form Pt silicides in the gaps. In another embodiment the semiconductor material is exposed to W in a silane mixture and the W replaces the semiconductor material so that the W projects into the material.
044252951	summary	The present invention relates to systems for achieving nuclear fusion. A number of writings are listed in this paragraph to serve as background for the explanation hereinafter, the writings listed here being merely representative: "Confining a Tokamak Plasma with rf-Driven Currents" (Fisch), Physical Review Letters, Vol. 41, Sept. 25, 1978, p. 873 (called Fisch (1978) herein); "System and Method for Generating Steady State Confining Current for a Toroidal Plasma Fusion Reactor" (Fisch), U.S. patent application, Ser. No. 935,222, filed Aug. 21, 1978 (called Fisch (1978b) herein); "Methods of Driving Current by Heating a Toroidal Plasma" (Fisch), Proceedings of the Second Joint Varenna-Grenoble International Symposium on Heating in Toroidal Plasma, Como, Italy, Sept. 3, 1980 (called Fisch (1980) herein); "Creating an Asymmetric Plasma Resistivity with Waves" (Fisch et al.), Physical Review Letters, Vol. 45, Sept. 1, 1980, p. 720 (called Fisch et al. (1980) herein); "Current Generation in a Relativistic Plasma" (Fisch), Princeton University Plasma Physics Laboratory Report PPPL-1763, January, 1981 (called "Fisch (1981)" herein); "Tokamak Research" (Furth), Nuclear Fusion Vol. 15 (1975) p. 487 (called "Furth (1975)" herein); "Theory of Electron Cyclotron Resonance Heating of Tokamak Plasmas" (Ott et al.), Physics of Fluids Vol. 23, May, 1980, p. 1031 (called Ott et al. (1980) herein); "New Methods of Driving Current in Fusion Devices" (Ohkawa), Nuclear Fusion, Vol. 10, (1970), p. 185 (called "Ohkawa (1970)" herein); "Steady-State Operation of Tokamaks by r-f Heating" (Ohkawa), General Atomic Report GA-A13847, Feb. 23, 1976 (called "Ohkawa (1976)" herein); "The Peristaltic Tokamak" (Wort), Plasma Physics, Vol. 13, 1971, p. 258 (called "Wort (1971)" herein). See, also, Coppi et al. U.S. Pat. No. 3,778,343. The operation of a tokamak is dependent upon the maintenance of a toroidal electric current to confine the plasma. For a fusion reactor based upon the tokamak concept to become an economic reality, this toroidal current must be produced both cheaply and in long pulses. Long pulses are required in order to limit the metal fatigue arising from the heat stress to which the structural components of the tokamak are subjected in a pulsed device. The method originally envisioned for driving this toroidal current is by means of a time-varying magnetic field which induces a toroidal electric field. This method suffers, however, in that it is inherently a pulsed method. In contrast, the invention described herein provides means of generating this current continuously. Moreover, to sustain this current in the manner prescribed by the present invention requires an amount of power that is small enough for the system to be extremely attractive in fusion applications. Accordingly, it is an object of the present invention to provide a system of steady-state toroidal electric currents in the plasma of a fusion device serving to confine the plasma. The present invention exploits in a novel manner the principle that the rate of coulomb collisions between charged particles is a sensitive function of the relative speed of the colliding particles. Means are provided to selectively heat electrons traveling in one toroidal direction in order to assure that these (heated) electrons collide less with the plasma ions than do the unheated electrons traveling in the opposite direction. Consequently, the ions drag preferentially on the more collisional electrons, with the result that a current is generated with electrons, on average, flowing in the direction of the heated electrons and with ions, on average flowing in the opposite direction. This mechanism for generating current is especially attractive because it does not rely upon an external source of momentum in the direction that the current flows. For example, consider a plasma immersed in a steady magnetic field. To generate a current in the direction of the magnetic field (herein denoted as the parallel direction), it suffices merely to increase the cyclotron motion of selected electrons in this plasma. This cyclotron heating can be accomplished by launching various radio frequency (rf) waves into the plasma. That the waves need not have substantial parallel momentum allows the advantageous use of waves with parallel phase velocity greater than the speed of light, c. Such waves may be brought to the plasma by means of waveguides and injected into the plasma through conveniently small apertures. Waveguides are a particularly appealing construction for bringing energy into a tokamak; they can easily be interspersed between the toroidal magnets due to their relatively small apertures and they may be bent to suit engineering requirements.
062657232	summary	BACKGROUND OF THE INVENTION The present invention relates to a magnetic shield room and, more particularly, to a magnetic shield room having an opening portion through which wafers and the like are unloaded and loaded. Conventionally, as a unit for forming a resist pattern on a sample such as a wafer or mask, an EB (Electron Beam) exposure unit which draws a pattern on a photoresist by using an electron beam is used. An EB exposure unit of this type has a problem that the electron beam irradiation position varies by the influence of the external magnetic field to undesirably cause a distortion in the drawn pattern. In order to prevent this, such an EB exposure unit is arranged in a shield room to operate while it is shielded from the external magnetic field. Concerning a wafer or mask to be processed by the EB exposure unit, a cassette which stores a plurality of wafers or masks is usually handled as one unit. Therefore, the EB exposure unit has a loading portion for extracting a wafer or mask as a processing target from the cassette and supplying it to the processing chamber, and an unloading portion for storing a processed wafer or mask in the cassette. Also, a means for attaching and detaching the cassette to and from the loading and unloading portions is necessary. When such loading and unloading portions are arranged in the magnetic shield room, conventionally, the cassette is manually loaded and unloaded by the operator or the like through an inlet/outlet port formed in the magnetic shield room and provided with a normally closed door. When the door is opened to allow the operator to enter or leave the room, the EB exposure unit is influenced by the external magnetic field. Therefore, when loading/unloading the cassette, operation of the EB exposure unit must be temporarily stopped, leading to a decrease in throughput. FIG. 6 shows a conventional magnetic shield room. As shown in FIG. 6, a maintenance door 3 is provided to the side wall of a magnetic shield room 1, and a magnetic field shield material is adhered to the inner wall of the magnetic shield room 1. An EB exposure unit (not shown) or the like is arranged in the magnetic shield room 1. With this arrangement, to prevent the influence of the external magnetic field, an opening portion having such a size that it does not allow the external magnetic field to influence the EB exposure unit may be formed in the magnetic shield room 1, and the cassette may be loaded/unloaded through this opening portion. Based on the demands for a higher micropatterning degree and a higher integration degree in recent semiconductor integrated circuits, a strict pattern drawing precision of 0.1 .mu.m or less has been required, and the magnetic field around the EB exposure unit must be suppressed as low as possible. If an opening portion having a size required for loading/unloading a cassette (e.g., one having a size of 180 mm.times.180 mm.times.180 mm) is formed, the external magnetic shield enters the magnetic shield room 1 through the opening portion to adversely affect the EB exposure unit. Therefore, the EB exposure unit must be installed to be sufficiently remote from the opening. As a result, the area occupied by the magnetic shield room 1 with respect to the area occupied by the EB exposure unit becomes considerably large. In order to set the strength of the entering external magnetic field not to influence the EB exposure unit, the length of the short sides of the rectangular opening portion must be decreased to 100 mm or less. With this size, however, at most only one wafer or mask can be passed through this opening portion, and wafers and masks stored in a cassette cannot be loaded/unloaded at all. As a method of suppressing entrance of the external magnetic field into the magnetic shield room 1 through the opening, one in which a tubular magnetic field shield material is provided to the outside of the opening portion is proposed, as shown in Japanese Patent Laid-Open No. 59-197198. In order to improve the magnetic shield effect without decreasing the opening ratio, a technique as shown in Japanese Utility Model Laid-Open No. 3-12497 is proposed, in which a stereoscopic shield lattice having a depth is further arranged in a tubular shield material arranged outside the opening portion, such that the interval pitch is smaller than the depth. In order to sufficiently shield the external magnetic shield by providing a tube made of a magnetic shield material at the opening portion, the length of the tubular member must be increased in accordance with the size of the opening. If, however, the tubular member is long, it interferes with the operability of loading/unloading the cassette in/from the loading and unloading portions in the magnetic shield room. For example, when the operability of placing the cassette on the loading or unloading portion is considered, the length of the tubular member is preferably as small as possible. When a shield lattice is arranged in the tubular member, it is suitable for an application such as a vent port. For an application, e.g., a case that includes loading/unloading of a cassette, the lattice interval must be increased. For this reason, the length of the tubular member must be increased in accordance with the size of the opening, thus interfering with the operability. SUMMARY OF THE INVENTION It is an object of the present invention to provide a magnetic shield room which suppresses entrance of an external magnetic field and allows loading/unloading of a cassette. It is another object of the present invention to provide a magnetic shield room in which operability of loading/unloading the cassette is improved. In order to achieve the above objects, according to the present invention, there is provided a magnetic shield apparatus comprising a magnetic shield room having an opening to shield external magnetism, a tubular member made of a magnetic shield material and attached to the opening to project from the magnetic shield room by a first predetermined length, and a flange portion made of a magnetic shield material and formed around a distal end portion of the tubular member to be spaced apart therefrom by a second predetermined length.
	abstract	An electrolytic apparatus for an oxide electrolytic method includes an interior of an electrolytic vessel, a common cathode and two types of anodes different in shape and arrangement, a first electrolysis controller is connected between the cathode and the first anode, and a second electrolysis controller is connected between the cathode and the second anode. The electrolytic processing of the substance in the electrolytic vessel is carried out such that a pair of the cathode and one of the anodes is used for main electrolysis and a pair of the cathode and the other anode is used for auxiliary electrolysis. By this apparatus, prevention of the ununiform distribution of the electrodeposit, improvement of the processing speed and improvement of the durability of the crucible are achieved, whereby the recycling of spent nuclear fuels based on the nonaqueous reprocessing method is made feasible in a commercial scale.
	description	The present application claims priority to U.S. provisional Application No. 61/041,266, filed on Apr. 1, 2008. 1. Field of the Invention The present invention relates to digital radiography and, more specifically, to a method for the detection of collimator blades in digital radiography images. 2. Description of the Related Art An essential step in processing digital radiography images is to detect the collimator blades. The information obtained from the detection of the collimator blades is then used to determine the area enclosed by the blades, and the statistics associated with the image within this area are calculated for use in subsequent image processing steps. The portion of image that is outside the area of the collimator blades may then be discarded to facilitate only useful image data storage, transmission and processing. This technique of identifying the target area and discarded undesired areas is commonly referred to as the Auto Shutter process. The area enclosed by the collimator blades is referred to as the shutter area. The typical Auto Shutter process comprises the two steps: (1) the use of edge detection algorithms, such as the Hough transform, to detect the potential edges of the collimator blades in the image, which appear as straight lines; and (2) the selection of desirable edges. Unfortunately, the Hough transform is often unable to detect all the desirable edges because some or all of the collimator blade edges may be too weak for successful detection. As a result, an erroneous area may be selected and the diagnostic quality of image will therefore be compromised. It is therefore a principal object and advantage of the present invention to provide a method for more accurately detecting collimator blade edges. It is an additional object and advantage of the present invention to provide a method for improving diagnostic image quality. It is a further object and advantage of the present invention to provide a method for improving useful image data storage, transmission and processing. In accordance with the foregoing objects and advantages, the present invention provides a process for detecting the edges of collimator blades in a digital radiography image that comprises at least two passes for improved edge detection and location of a target area in the image. The first pass in the process is to use the captured image (i.e., original image) to detect the edges of the collimator blades. This pass may be performed by implementing conventional edge detection processes and algorithms, such as the Hough transform. The second pass in the process is to repeat edge detection using an enhancing image. Image enhancement may be accomplished by using a histogram matching technique. This pass may also be repeated any number of times in cases of complex anatomy or when selected radiographic techniques do not provide sufficient imaging data. The results of the second pass, or the collection of the results of multiple second passes, are then combined with the result from the first pass to form a list of the potential blade edge candidates. A desirable number of edges are then selected from the combined list to form a polygon which encloses the target area of the image, thereby providing the shutter area. Referring now to the drawings, wherein like reference numerals refer to like parts throughout, there is seen in FIG. 1 a conventional method for detecting collimator blades. In the conventional method, an original image is processed using edge detection to detect a list of potential edges of collimator blades in the image, which appear as straight lines. A sample number of edges, such as four to eight, and then selected and used to determine the shutter configuration. Referring to FIG. 2, there is seen a method 10 for detecting the collimator blade edges in the digital radiography image according to the present invention. Method 10 begins with the original image 12. Image 12 is analyzed in a first edge detection pass 14 to detect the edges of any collimator blades within the image. Referring to FIG. 2, step 14 may be performed by implementing conventional edge detection algorithms, such as the Hough transform, on the digital image. The detected edges are then stored in first edge list 16. The original image 12 is enhanced 18 resulting in an enhanced image 20. This enhanced image is analyzed in a second edge detection pass 22. The detected edges are stored in a second edge list 24. The edges in first edge list 16 and second edge list 24 are combined to forma a combined edge list 26. All the edges in the combined edge list 26 are then validated 28, and a predetermined or desirable number of edges are selected as representing the collimator blades. Finally, a polygon is drawn 30 to enclose the shutter area. In method 10, as seen in FIG. 2, the left branch containing first edge detection pass 16 has the same functionality as the counterpart in the conventional method, as seen in FIG. 1. However, the right branch of method 10 containing second edge detection pass 22 is designed particularly to detect weak edges in the enhanced image. This processing may be repeated any number of times in cases of complex anatomy or when selected radiographic techniques does do not provide sufficient imaging data. Image enhancement 18 is accomplished using the histogram matching technique. In brief, the enhanced image is generated by modifying the pixel values of the original image in such a way that the histogram of the original image (“source histogram”) is modified to match the “destination histogram” of the enhanced image. The “destination histogram” is so designed as to raise weak edges in the image. There is seen in FIG. 3 an enhanced image 20 including a polygon 32 to enclose the shutter area 34. It should be recognized by those of skill in the art that once the shutter area 30 is defined, the image may be cropped or otherwise handled according to conventional methods or devices, such as data storage, transmission and processing.
	claims	1. An ion implantation system comprising:means for generating an ion beam; means for determining an ion beam current reference level; means for measuring an ion beam current during implantation; and means for adjusting an ion implantation parameter to compensate for vacuum fluctuations during implantation based on the reference level and the measured ion beam current, and not based on a detected pressure. 2. An ion implantation system comprising:a beam generator that generates an energetic ion beam and directs the beam along an ion beam path toward a semiconductor wafer; a detector that detects an ion beam current; a wafer drive that moves the semiconductor wafer in a direction transverse to the ion beam path; and a controller that receives signals from the detector representative of a detected ion beam current, detects a vacuum fluctuation based on the a difference value determined from an ion beam current reference value, which corresponds to an ion beam current in the absence of vacuum fluctuations along the ion beam path, and a detected ion beam current measured in the presence of vacuum fluctuations along the ion beam path, and controls the wafer drive to adjust a wafer scan rate to compensate for the vacuum fluctuation during implantation. 3. The apparatus of claim 2, wherein the controller scales the difference value to account for non-line of sight and line of sight charge exchanging collisions experienced by ions in the beam along the ion beam path. 4. The apparatus of claim 3, wherein the difference value is scaled based on a ratio of line of sight collisions to non-line of sight collisions. 5. The apparatus of claim 2, further comprising a vacuum system, and wherein the controller controls the vacuum system to begin evacuation based on the determined difference value. 6. The apparatus of claim 2, wherein the detector is a Faraday cup positioned adjacent a semiconductor wafer. 7. The apparatus of claim 2, wherein the beam generator includes an angle corrector magnet. 8. The apparatus of claim 2, wherein the ion beam current reference value is determined based on an ion beam current measured while a vacuum level along the ion beam path is stable. 9. The apparatus of claim 2, wherein the ion beam current reference value is retrieved by the controller from a memory. 10. The apparatus of claim 2, wherein the controller detects a vacuum fluctuation based on a difference value between an ion beam current reference value, which corresponds to an ion beam current in the absence of vacuum fluctuations along an ion beam path, and an ion beam current measured in the presence of vacuum fluctuations along the ion beam path. 11. The apparatus of claim 2, wherein the controller adjusts an ion implantation parameter in addition to the wafer scan rate to adjust for wafer dosing non-uniformity in two dimensions. 12. The apparatus of claim 2, wherein the controller adjusts a wafer scan rate and a beam scan rate. 13. The apparatus of claim 12, wherein the controller adjusts the wafer scan rate and beam scan rate based on two scale factors. 14. The apparatus of claim 2, wherein the controller adjusts the wafer scan rate using a scale factor that is mathematically derived by modeling the implantation system. 15. The apparatus of claim 14, wherein the controller uses a scale factor that has been determined based on calculated beam path length neutral particle density products that are obtained, at least in part, from a model of an ion beam path and a vacuum system in the implantation system. 16. An ion implantation system comprising:a beam generator that generates an energetic ion beam and directs the ion beam toward a semiconductor workpiece; a detector that detects an ion beam current; and a controller that receives signals from the detector representative of a detected ion beam current, and controls at least one ion implantation parameter to compensate for vacuum fluctuation during implantation based on a difference value determined from an ion beam current reference value, which corresponds to an ion beam current in the absence of vacuum fluctuations along an ion beam path, and the detected ion beam current. 17. The system of claim 16, wherein the controller controls the at least one ion implantation parameter based on the difference value and not based on a detected pressure. 18. The system of claim 16, wherein the controller scales the difference value to account for non-line of sight and line of sight charge exchanging collisions experienced by ions in the ion beam along the ion beam path. 19. The system of claim 18, wherein the difference value is scaled based on a ratio of line of sight collisions to non-line of sight collisions. 20. The system of claim 16, further comprising a vacuum system, and wherein the controller controls the vacuum system to begin evacuation based on the determined difference value. 21. The system of claim 16, wherein the detector is a Faraday cup positioned adjacent a semiconductor wafer. 22. The system of claim 16, wherein the beam generator includes an angle corrector magnet. 23. The system of claim 16, wherein the ion beam current reference value is determined based on an ion beam current measured while a vacuum level along the ion beam path is stable. 24. The system of claim 16, wherein the ion beam current reference value is retrieved by the controller from a memory. 25. The system of claim 16, wherein the controller adjusts an ion implantation parameter to adjust for semiconductor workpiece dosing non-uniformity in two dimensions. 26. The system of claim 16, wherein the at least one ion implantation parameter includes one of a wafer scan rate and a beam scan rate. 27. The system of claim 16, wherein the controller determines an adjusted difference value using a scale factor and the difference value, and uses the adjusted difference value to control the at least one ion implantation parameter. 28. The system of claim 16, wherein the controller controls the at least one ion implantation parameter based on the difference value and a scale factor that is mathematically derived by modeling the implantation system. 29. The system of claim 28, wherein the controller uses a scale factor that has been determined based on calculated beam path lengthneutral particle density products that are obtained, at least in part, from a model of an ion beam path and a vacuum system in the implantation system. 30. An ion implantation system comprising:a beam generator that generates an energetic ion beam and directs the ion beam along an ion beam path toward a semiconductor workpiece, the ion beam path being non-linear; a detector that detects an ion beam current; and a controller that receives signals from the detector representative of a detected ion beam current, and controls at least one ion implantation parameter based on the detected ion beam current and a ratio of line of sight to non-line of sight collisions between the particles in the ion beam and other particles along the ion beam path to compensate for vacuum fluctuation during implantation. 31. The system of claim 1, wherein the means for adjusting determines a difference value between the ion beam current reference value, which corresponds to an ion beam current in the absence of vacuum fluctuations along an ion beam path, and the measured ion beam current. 32. The system of claim 31, wherein the means for adjusting scales the difference value to account for non-line of sight and line of sight charge extending collisions experienced by ions in the ion beam along the ion beam path. 33. The system of claim 31, wherein the means for adjusting controls the at least one ion implantation parameter based on the difference value and a scale factor that is mathematically derived by modeling at least a portion of the implantation system. 34. The system of claim 31, wherein the means for adjusting uses a scale factor that has been determined based on calculated beam path lengthneutral particle density products that are obtained, at least in part, from a model of an ion beam path and a vacuum system in the implantation system. 35. The system of claim 31, wherein the means for adjusting adjusts the ion implantation parameter based on a ratio of line of sight collisions to non-line of sight collisions experienced by ions in the ion beam along the ion beam path. 36. The system of claim 1, further comprising a vacuum system, and wherein the means for adjusting controls the vacuum system to begin evacuation based on the determined difference value. 37. The system of claim 1, wherein the means for measuring includes a Faraday cup positioned adjacent a semiconductor workpiece. 38. The system of claim 1, wherein the means for generating includes an angle corrector magnet. 39. The system of claim 1, wherein the ion beam current reference value is determined based on an ion beam current measured while a vacuum level along an ion beam path is stable. 40. The system of claim 1, wherein the means for determining retrieves the ion beam current reference value from a memory. 41. The system of claim 1, wherein the means for adjusting detects a vacuum fluctuation based on a difference value determined from an ion beam current reference value, which is an ion beam current measured in the absence of vacuum fluctuations along an ion beam path, and an ion beam current measured in the presence of vacuum fluctuations along the ion beam path. 42. The system of claim 1, wherein the means for adjusting adjusts an ion implantation parameter to adjust for wafer dosing non-uniformity in two dimensions. 43. The system of claim 1, wherein the at least one ion implantation parameter includes one of a wafer scan rate and a beam scan rate.
047434244	summary	BACKGROUND OF THE INVENTION Field of the Invention The invention concerns a nuclear reactor installation comprising a cavity which is lined with a liner and arranged in a prestressed concrete pressure vessel. A high temperature reactor is provided on a circular base plate in the cavity. A plurality of core rods may be inserted into and retracted from an internal space containing spherical fuel elements by means of rod drives; and a plurality of reflector rods may be inserted into and retracted from a side reflector of said reactor by means of rod drives. The spherical fuel elements may be introduced by means of fuel element feeder-tubes and removed by at least one vertical, cooled outlet tube disposed on a base reflector. The nuclear reactor installation also has an annular cylindrical interstitial space formed between a thermal side shield and the liner. The thermal side shield surrounds a side reflector which forms a reactor core. The thermal side shield supports said reflector with radial supports. First heat exchangers, preferably steam generators, are arranged uprightly, aligned with an axial first passage through the roof of the pressure vessel. Said first heat exchangers are also interconnected with first blowers of a gas circulation loop through the nuclear reactor. Nuclear reactor installations of this type must satisfy particularly high safety requirements. This leads, in the case of known installations, particularly those with high capacity, to very expensive and cost intensive structures. Such added expense holds true for both the prestressed concrete pressure vessel and the development and arrangement of the individual structural elements of the nuclear reactor. SUMMARY OF THE INVENTION It is therefore an object of the invention to provide a nuclear reactor of the aforementioned type, particularly one with a capacity of approx. 300 MWe to 600 MWe; one which is suitable both for generating steam and supplying process heat; and one which has a simplified construction or individual structural elements, respectively, which are simply designed. It is also an object of the present invention to provide a nuclear reactor which allows radial expansions in a simple manner. It is a further object of the present invention to provide a nuclear reactor with a uniform capacity distribution of the spherical fuel elements so that hot strands are avoided. It is an object of the present invention to provide a nuclear reactor with simplified rod drive mechanisms and corresponding control means. It is an object of the present invention to provide a nuclear reactor which ensures a uniform passage rate of the fuel elements through the reactor core so as to increase efficiency. It is also an object of the present invention to provide a nuclear reactor in which radiation induced stresses in the reflector rods are reduced and limited. All of these objects should be met by the nuclear reactor according to the invention without comprising the reactors satisfaction of safety requirements. The above objects of the invention are attained with a nuclear reactor plant of the aforementioned type wherein: the bottom plate is composed of segments which are centered around a central reference point, and which are supported by means of radially displaceable double supports on the bottom of the pressure vessel; core rods are equipped with rod drives by means of which the core rods may be inserted, in a concerted manner, to only a part of the layer height of the spherical fuel elements, i.e., all the rods move simultaneously and to the same height; the reflector rods are equipped with electric motor rod drives; several uniformly distributed fuel element outlet tubes are provided in the bottom reflector for the discharge of fuel elements which pass through the nuclear reactor; first blowers are arranged in first passages through the roof of the pressure vessel, and disposed on slides guided in magnetic bearings; at least one upright heat exchanger is provided in the interstital space for removing the decay heat; a second passage, with which said heat exchanger is aligned, passes lengthwise through the bottom of the pressure vessel; a corresponding second blower is provided in said second passage, also dispositioned on slides with magnetic bearings; radial supports, particularly in the case of nuclear reactors with a capacity in excess of 400 MVe, are designed to be elastic and to prevent relative movements between the side reflector and the thermal side shield and are provided in at least the upper third of the height of the side reflector. Further objects, features, and advantages of the present invention will become apparent from the detailed description of preferred embodiments which follows, when considered together with the attached figures of drawings.
054486122	abstract	An X-ray apparatus is disclosed which includes a mirror having a reflection surface, for expanding an X-ray beam in a predetermined direction, a detecting device for detecting a relative positional relationship between the X-ray beam and the reflection surface with respect to a direction perpendicular to the reflection surface, and an adjusting device for adjusting the relative position of the X-ray beam and the reflection surface on the basis of the detection. Also disclosed is an exposure apparatus and a semiconductor device manufacturing method using the X-ray apparatus.
045377405	summary	BACKGROUND OF THE INVENTION This invention relates to systems used to detect and locate failures of the cladding of fuel rods in nuclear fuel assemblies used in nuclear reactors, especially sodium cooled nuclear reactors. Nuclear reactors contain a fuel core which is a grouping of fuel assemblies each of which has a plurality of fuel rods. A fuel rod is a cylindrical, metal tube which contains nuclear fuel pellets. The metal of the tube separates the fuel pellets from reactor coolant which flows over the surface of the tube or cladding. A penetration of the cladding, termed a fuel failure, may allow fission fragments, particularly gases, to escape from the fuel rod into the reactor coolant. These gases may mingle with the gas contained in a cover gas region which is usually a feature of liquid metal cooled reactors. Nuclear reactors are expected to experience fuel failure in spite of rigorous quality control and conservative operating procedures. Most of the failures result from pin-hole cracks in the cladding and/or end plug welds. Such failures are now detected by analysis of fission-gas outside the core (e.g., in the reactor cover gas) and by observation of delayed neutron precursors in the reactor coolant. One of the problems faced by reactor instrumentation is to detect and monitor failed fuel in such a manner that safe operation of the reactor is not impaired. This problem can be solved by having the ability to quickly obtain and analyze samples of fission gas released by failed fuel. There is a major need to locate the leaking fuel assembly rapidly in order to expedite its removal and minimize reactor down time. Consequently, it is desired to provide a method for obtaining a sample of fission gas from reactor coolant and quickly identify the leaking assembly. SUMMARY OF THE INVENTION A sample of fission gas which may have been released by fuel pins into the reactor coolant via cladding defects is obtained as per this invention by passing a portion or all of the reactor coolant flow through a filter having holes therein of size appropriate in consideration of the pressure drop across the filter, to cause bubbles of fission gas entrained within the reactor coolant to be restrained from passage through the filter due to the surface tension forces on the bubble.
	claims	1. An apparatus for inspecting welds in a nuclear reactor, comprising:a body;a rotatable pad on the body;a first horizontal pad and a second horizontal pad on opposing sides of the body;a first shaft and a second shaft connected to the body, the first and the second horizontal pads being slideably mounted on the first shaft and the second shaft, respectively, to configure the body to move in a vertical direction;a first vertical pad and a second vertical pad on opposing sides of the body;a third shaft and a fourth shaft connected to the body, the first vertical pad and the second vertical pad being slideably mounted on the third shaft and the fourth shaft, respectively, to configure the body to move in a horizontal direction; andan inspection device. 2. The apparatus of claim 1, wherein the rotatable pad is located at a central portion of the body to rotate the inspection device. 3. The apparatus of claim 2, wherein a rotation of the rotatable pad is in increments of 90 degrees. 4. The apparatus of claim 1, further comprising:legs, each of the legs being attached to a respective side of the body, each of the first shaft, the second shaft, the third shaft and the fourth shaft being held by a respective one of the legs. 5. The apparatus of claim 4, wherein each of the legs is attached to the body by a stem member. 6. The apparatus of claim 4, wherein each of the legs is substantially Y-shaped or U-shaped. 7. The apparatus of claim 1, further comprising:a forked arm on each side of the body,wherein each of the first horizontal pad and the second horizontal pad, and the first vertical pad and the second vertical pad, being mounted on one of the respective forked arms. 8. The apparatus of claim 7, wherein each of the forked arms is substantially Y-shaped or U-shaped. 9. The apparatus of claim 7, further comprising:a support member attached to each of the forked arms,wherein each of the support members includes a hole for inserting one of the first shaft, the second shaft, the third shaft and the fourth shaft, respectively, each of the respective support members being configured to slideably move the first horizontal pad, the second horizontal pad, the first vertical pad and the second vertical pad, respectively, along the first shaft, the second shaft, the third shaft and the fourth shaft, to move the body in the respective horizontal or vertical directions. 10. The apparatus of claim 9, wherein each of the first horizontal pad, the second horizontal pad, the first vertical pad and the second vertical pad each respectively slide in an axial direction along the first shaft, the second shaft, the third shaft and the fourth shaft. 11. The apparatus of claim 9, wherein the first shaft, the second shaft, the third shaft and the fourth shaft each are a ball screws that translate rotational motion to linear motion to configure the body to move in the respective horizontal or vertical direction. 12. The apparatus of claim 4, wherein the inspection device is mounted to one of the legs. 13. The apparatus of claim 12, wherein the inspection device is an ultrasonic probe. 14. The apparatus of claim 13, wherein the ultrasonic probe is attached to a gimbal sensor. 15. The apparatus of claim 12, wherein the inspection device is supported by support arms. 16. The apparatus of claim 1, further comprising:a vacuum system to controllably adhere or force the first horizontal pad, the second horizontal pad, the first vertical pad and the second vertical pad to a surface of a core shroud. 17. The apparatus of claim 1, wherein the first shaft and the second shaft are about parallel to each other, and the third shaft and the fourth shaft are about parallel to each other. 18. The apparatus of claim 16, wherein the vacuum system includes,a venturi valve,a hose connected to the venturi valve, anda pump in fluid communication with the hose and the venturi valve, the pump and the venturi valve being configured to create a vacuum force to adhere the first horizontal pad, the second horizontal pad, the first vertical pad and the second vertical pad to the surface of the core shroud.
040630986	claims	1. A system for deflecting an input beam of charged particles having a spread of momenta from a source thereof to cause said beam to impinge on a target having a predetermined spatial relationship with said source, said system comprising means responsive to said input beam for periodically scanning said beam in a scanning plane; means responsive to said periodically scanned beam for deflecting said periodically scanned beam through preselected angle in a deflection plane, the said scanning plane subsequent to said deflection being angularly oriented with respect to said scanning plane prior to said deflection, said deflected and periodically scanned beam being thereupon directed to impinge upon said target plane; and means for substantially reducing the momentum dispersion of said beam in the deflection plane at said target due to the deflection thereof by said deflecting means and for substantially reducing the momentum dispersion of said beam in the scanning plane at said target due to the periodic scanning movement thereof by said beam scanning means. means for spatially focusing said beam to produce a beam diameter of a predetermined size at said target. 2. A system in accordance with claim 1 and further including 3. A system in accordance with claim 1 wherein said deflection and scanning dispersion reducing means includes a quadrupole magnet positioned in the path of said beam following said deflecting means. 4. A system in accordance with claim 3 wherein said quadrupole magnet has an asymmetric configuration for reducing the dispersion of said beam in the scanning plane due to the scanning movement thereof. 5. A system in accordance with claim 4 wherein the asymmetry of said quadrupole magnet is arranged to provide a first magnetic field strength in the scanning plane of said beam for deflecting the particles of said beam having lower momenta and a second magnetic field strength in the scanning plane of said beam for deflecting the particles of said beam having higher momenta, thereby reducing the dispersion of the beam in the scanning plane at said target. 6. A system in accordance with claim 5 wherein the asymmetry of said quadrupole magnet is arranged to provide an effective sextupole component of magnetic field strength for reducing the dispersion in said scanning plane superimposed on the quadrupole component of magnetic field strength which reduces dispersion in the deflecting plane. 7. A system in accordance with claim 6 wherein said deflection means is a dipole magnet. 8. A system in accordance with claim 2 wherein said deflection and scanning dispersion reducing means includes a quadrupole magnet positioned in the path of said beam following said deflecting means. 9. A system in accordance with claim 8 wherein said quadrupole magnet has an asymmetric configuration for reducing the dispersion of said beam in the scanning plane due to the scanning movement thereof. 10. A system in accordance with claim 9 wherein the assymmetry of said quadrupole magnet is arranged to provide a first magnetic field strength in the scanning plane of said beam for deflecting the particles of said beam having lower momenta and a second magnetic field strength in the scanning plane of said beam for deflecting the particles of said beam having higher momenta, thereby reducing the dispersion of the beam in the scanning plane at said target. 11. A system in accordance with claim 10 wherein the asymmetry of said quadrupole magnet is arranged to provide an effective sextupole component of magnetic field strength for reducing the dispersion in said scanning plane superimposed on the quadrupole component of magnetic field strength which reduces dispersion in the deflecting plane. 12. A system in accordance with claim 11 wherein said deflection means is a dipole magnet and said spatial focusing means includes the arrangement of the entrance and exit angles of said deflection dipole magnet for producing desirable focusing effects on said beam in the deflection plane and the scanning plane thereof. 13. A system in accordance with claim 12 wherein said spatial focusing means further includes a pair of quadrupole magnets positioned in the path of said beam ahead of said beam scanning means, the magnetic field strengths of each of said quadrupole magnets being selected to produce focusing effects on said beam in the deflection plane and the scanning plane thereof.
	claims	1. A structure for preventing a scan by a beam, the structure comprising:a primary material forming the structure, the primary material comprising a first mass attenuation coefficient enabling the primary material to be penetrated by the beam; anda matrix of particles within the primary material to provide scattering or attenuating of the beam to distort the scan,wherein the particles comprise one or more secondary materials different than the primary material,wherein the one or more secondary materials comprises a plurality of crystal particles distributed in three-dimensional modified matrix with a varying number of the plurality of crystal particles located in offset positions and with a varying number of grouped particles, a subset of the plurality of crystal particles comprising oblong shaped crystal particles, and a second subset of the plurality of crystal particles comprises round sphered crystal particles,wherein the one or more secondary materials comprises at least one subsequent mass attenuation coefficient that is greater than the first mass attenuation coefficient of the primary material, andwherein the at least one subsequent mass attenuation coefficient enables the particles to scatter or attenuate the beam to distort the scan comprising the varying number of the grouped particles being positioned within the structure to prevent a view of a design feature or internal component to the structure by the scan,wherein the matrix of particles comprises one or more gaps to enable geometric dimensioning and tolerancing measurements and inspection of critical areas of the structure, one or more vacant areas that include no particles to reveal a view of a first design feature of the structure, one or more secondary materials located in at least one cluster implemented to distort a view of a second design feature of the structure, and one or more vacant areas that include no particles to mislead a scan and analysis of the view of the first and second design features. 2. The structure of claim 1, wherein the primary material comprises aluminum and the one or more secondary materials comprises tungsten, copper, nickel, or iron. 3. The structure of claim 1, wherein the matrix of particles is uniform. 4. The structure of claim 1, wherein the matrix of particles comprises one or more secondary materials located in offset positions. 5. The structure of claim 1 comprises a component, a part, or a tool utilized in an electro-mechanical system of an aircraft. 6. The structure of claim 1, wherein the primary material is layered via additive manufacturing technologies to form the structure. 7. The structure of claim 1, wherein the primary material is produced via casting technologies to form the structure.
046474218	abstract	An operation control method for a nuclear reactor performing a load follow-up operation in accordance with a load variation program, wherein a core reactivity which changes with time in a first cycle of operation is predicted based on the load variation program and data for analyzing dynamic characteristics of the reactor, and a change in a liquid poison concentration in the first cycle of operation is obtained based on the predicted reactivity, and reactor power is controlled in the first cycle of operation by adjusting the liquid poison concentration in accordance with the obtained change. When the liquid poison concentration is adjusted and control rods are manipulated, a reactivity introduced by these operations in the first cycle of operation is obtained, and an adjustment to be made to the liquid poison concentration in the second cycle of operation which is the next cycle of operation following the first cycle of operation is obtained from the reactivity introduced in the first cycle of operation. When the change in reactivity in the second cycle of operation becomes equal to that in the first cycle of operation, and reactor power is controlled in the second cycle of operation by performing the obtained adjustment of the liquid poison concentration.
	summary
063109387	abstract	The present invention is, in one embodiment, a method for determining tracking control parameters for positioning an x-ray beam of a computed tomography imaging system having a movable collimator positionable in steps and a detector array including a plurality of rows of detector elements. The method includes steps of obtaining detector samples at a series of collimator step positions while determining a position of a focal spot of the x-ray beam; determining a beam position for each detector element at each collimator step utilizing the determined focal spot positions, a nominal focal spot length, and geometric parameters of the x-ray beam, collimator, and detector array; and determining a calibration parameter utilizing information so obtained. For example, in determining a target beam position at which to maintain the x-ray beam, a detector element differential error is determined according to ratios of successive collimator step positions; and a target beam position is selected for an isocenter element in accordance with the determined element differential errors.
052710513	claims	1. In a nuclear reactor comprising a reactor vessel having a reactor core fueled with fuel assemblies, a reactor coolant system circulating coolant through the reactor core, a refueling cavity through which fuel assemblies are inserted into and removed from the reactor core, a spent fuel pit in which fuel assemblies removed from the reactor core are stored under coolant, fuel transfer means through which fuel assemblies are transferred between said refueling cavity and said spent fuel pit under coolant, a refueling water storage tank containing a supply of coolant which fills said refueling cavity during refueling, spent fuel pit cooling means, skimmer means for skimming particulates from the coolant, and purification means for purifying said coolant, the improvement: wherein said spent fuel pit has a volume of coolant for removing decay heat from said fuel assemblies in said spent fuel pit without operation of said spent fuel pit cooling means; wherein said spent fuel pit cooling means is non-safety rated; and wherein said spent fuel pit cooling means and said purification means are connected in series to form a combined cooling and purification system having common non-safety rated pump means and first piping means for circulating coolant drawn from said spent fuel pit through said combined cooling and purification system and back to said spent fuel pit to simultaneously in a single loop cool and purify spent fuel pit coolant; and including second piping means for selectively circulating coolant drawn from said refueling cavity through said combined cooling and purification system and back to said refueling cavity to simultaneously cool and purify refueling coolant in said cavity. 2. The improvement of claim 1 wherein said skimmer means is connected to said first piping means through which coolant is drawn from said spent fuel pit by said common pump means for circulation through said combined cooling and purification system. 3. The improvement of claim 1 wherein said skimmer means is connected to said second piping means through which coolant is drawn from said refueling cavity by said common pump means for circulation through said combined cooling and purification system. 4. The improvement of claim 1 including third piping means disposed between said refueling water storage tank and said second piping means for circulating coolant from said refueling water storage tank through said combined cooling and purification system and back to said refueling water storage tank using said common pump means. 5. The improvement of claim 4 wherein said second piping means and said third piping means have valves for circulating coolant from said refueling water storage tank through said combined cooling and purification system and directly to said refueling cavity without passing through said reactor core using said common pump means. 6. The improvement of claim 1 including transfer means selectively transferring coolant between said refueling cavity and said refueling water storage tank through said combined cooling and purification system without passing through said reactor core using said common pump means. 7. The improvement of claim 1 wherein said combined cooling and purification system comprises parallel branches each having cooling means, purification means and a common pump, and wherein said first piping means comprises common suction pipe means drawing coolant from said spent fuel pit for both branches, and common discharge pipe means discharging coolant from both branches into said spent fuel pit. 8. The improvement of claim 1 wherein said cooling means comprises heat exchanger means. 9. The Improvement of claim 1 wherein said purification means comprises demineralizer means. 10. The improvement of claim 9 wherein said purification means further includes filter means.
	abstract	A floating nuclear power reactor includes a self-cooling containment structure and an emergency heat exchange system. The containment structure of the reactor may be flooded upon the temperature or pressure in the containment structure reaching a certain level. The reactor vessel may also be flooded upon the temperature or pressure in the reactor vessel reaching a predetermined level. The reactor includes a heat exchange system and a filtered containment venting system.
	description	Referring now to the drawings, particularly to FIG. 1, there is illustrated a reactor pressure vessel, generally designated 10, having a reactor pressure vessel wall 12 and an inner core shroud 14 defining a generally annular space 16 therebetween. The annular space 16 contains coolant. As in a typical boiling water nuclear reactor, a plurality of jet pumps, one being generally designated 18, are disposed at circumferential spaced positions about the pressure vessel between the pressure vessel wall 12 and the core shroud 14 and in the annular space 16. Each jet pump 18 typically comprises an inlet riser 20, a transition piece 28 adjacent the upper end of the inlet riser 20, a pair of elbows 22, inlet-mixers 23, each including nozzles 24 and mixing sections 25, and diffusers 26. Holddown assemblies adjacent the top of the jet pump 18, together with a number of braces and restraints maintain each jet pump 18 in fixed position in the annular space 16 between the core shroud 14 and pressure vessel wall 12. A thermal sleeve 32 penetrates the pressure vessel wall 12 and is welded at its juncture with an inlet elbow. The opposite end of the inlet elbow is secured to the lower end of the inlet riser 20. It will be appreciated that the foregoing-described jet pump 18 is conventional in construction. Thus, coolant enters the thermal sleeve 32 and flows through the elbow, upwardly in the inlet riser 20, through the inlet elbows 22 through nozzles 24 for flow in a downward direction through the mixing sections 25, the diffusers 26 and into a plenum 40 for upward flow through the reactor core. As conventional, the jet pump nozzles 24 induce a suction flow of coolant from the annular space 16 into the mixing section 25 which mixes with the coolant flow through the jet pump nozzles 23. Referring more particularly to FIG. 2, there is illustrated a portion of a jet pump 18 having an inlet elbow 22 adjacent five nozzles 24. The nozzles 24 are supported above the mixing sections 25 and define therewith a generally annular suction flow passage 29 between the nozzles 24 and an inlet to the mixing section 25. It will be appreciated that the mixing section 25 is a cylindrical pipe which terminates at its lower end in an inlet to the diffuser 26. Consequently, the flow of coolant through the nozzles 24 induces a suction flow of coolant through the annular spacer 16 for flow into the mixing section 25. These combined nozzle and suction flows pass through the mixing section 25 and diffuser 26 and into plenum 40. Referring now to FIG. 3, there is illustrated two of the nozzles 24. It will be appreciated that the interior passages through nozzles 24 are conical in shape with the diameter decreasing along the path of the fluid flow, thereby increasing the flow velocity into mixing section 25. The increased velocity induces additional fluid to flow into the sleeve through the annular opening 29 between the nozzles 23 and the mixer sleeve inlet as indicated by the arrows in FIG. 2. In accordance with a preferred embodiment of the present invention, the inlet-mixer is provided with a coating that inhibits or eliminates xe2x80x9ccrudxe2x80x9d build-up. To accomplish this, the inlet-mixer 23 is placed in a chemical vapor deposition (xe2x80x9cCVDxe2x80x9d) reactor. The reactor is a heated vacuum vessel that is sufficiently large to house these parts. The vessel is then evacuated and the pressure is dropped to approximately 20 mtorr. Heat is applied to raise the temperature of the vessel and the part to a reaction temperature within a range of about 400xc2x0-500xc2x0 C. and preferably about 450xc2x0 C. When the vessel reaches the reaction temperature and pressure, chemical precursors, such as Ti(OC2H5)4 or Ta(OC2H5)5, are vaporized in the reactor chamber as a gas. These precursors impinge on the surface of the heated inlet-mixer part and thermally decompose to form a ceramic oxide coating, comprising, e.g., TiO2 or Ta2O5, and byproduct gases. The coating continues to form and to grow until the gas flow is terminated and the temperature decreased. When a sufficiently thick coating is achieved, e.g., within a range of about 0.5-3 microns and preferably about 1.0 micron, heating is terminated and the vessel cools. The vacuum is then released and the coated jet pump part removed. The coating is indicated 31 in FIGS. 2 and 3 along the interior wall surfaces of the inlet-mixer 23. The coating may comprise any dielectric coating, e.g., tantala (tantalum oxide, Ta2O5), titania (titanium oxide TiO2), and zirconia (ZrO2). However, in the preferred form, the dielectric coating is comprised of a ceramic oxide, preferably TiO2 or Ta2O5. Thus, the application of this ceramic oxide coating reduces the electrical potential between the metal of the inlet-mixers and the charged particles in the water, minimizing or eliminating the build-up of xe2x80x9ccrudxe2x80x9d on the surfaces of the inlet-mixers. That is, the rate of ion movement toward the inlet-mixer surface is significantly reduced or eliminated. Further, as a result of the above, the coating also serves to retard or eliminate stress corrosion cracking. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
046363360	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention broadly relates to drying an organic amine chelating agent and more particularly to the volume reduction of an aqueous medium containing the same. In one of its more particular aspects this invention relates to a process for reducing the volume of a low-level radioactive aqueous waste containing an organic amine chelating agent. In another of its more particular aspects, this invention relates to a process for producing a dry, flowable powder from such a waste. 2. Prior Art Waste management frequently involves the necessity of disposing of large volumes of materials, some of which may be contaminated with hazardous substances. In nuclear power plants, for example, large amounts of radioactive liquid and solid wastes are produced. Low-level radioactive wastes differ from high-level radioactive wastes, which are produced in the reprocessing of nuclear fuels, in that the latter present greater risks of contamination and therefore require disposal techniques which are more stringent than in the case of low-level radioactive wastes. Disposal of radioactive wastes in general cannot be readily accomplished by using conventional waste disposal techniques. Because of the relatively long half-lives of certain radioactive elements, the most widely used disposal techniques are storage, solidification and burial. The expense of so disposing of large volumes of radioactive wastes, however, is constantly rising and approaching levels at which volume reduction becomes not only economically desirable but a necessity. Many efforts have been directed at reducing the volume of radioactive wastes. U.S. Pat. No. 3,101,258 describes a heated-wall spray calcination reactor useful for disposing of nuclear reactor waste solutions. In spray calcination reactors of the heated-wall type, however, the temperature gradient from the outside of the reactor inward may result in uneven heating, producing regions of undesired high temperatures and causing non-uniform results. U.S. Pat. No. 3,922,974 discloses a hot air-fired furnace for incinerating radioactive wastes. The use of this apparatus, however, results in the production of noxious off-gases which require additional processing for removal. U.S. Pat. No. 4,145,396 describes a process for reducing the volume of organic waste material contaminated with at least one volatile compound-forming radioactive element selected from the group consisting of strontium, cesium, iodine and ruthenium. The selected element is fixed in an inert salt by introducing the organic waste and a source of oxygen into a molten salt bath maintained at an elevated temperature to produce solid and gaseous reaction products. The molten salt bath comprises one or more alkali metal carbonates and may optionally include from 1 to about 25 wt. % of an alkali metal sulfate. Although effective to some extent in reducing the volume or organic wastes, further volume reduction involving the separation of the radioactive materials from the non-radioactive components of the molten salt bath requires a number of additional processing steps. In U.S. patent application Ser. No. 451,516, filed Dec. 20, 1982 (now U.S. Pat. No. 4,499,833) and assigned to the assignee of the present invention, there is proposed a process for converting radioactive wastes in the form of liquids, solids and slurries into a mixture of a non-radioactive gas and a radioactive inorganic ash. In accordance with that process the radioactive waste is introduced as a finely atomized spray into a zone heated by means of a hot gas to a temperature sufficient to effect the desired conversion, preferably a temperature in the range of about 600.degree. to 850.degree. C. The process is conducted in a spray dryer modified to combust or calcine the waste. While the foregoing patent application discloses a process which is satisfactory for destroying most radioactive wastes, the high-temperature utilized in the process can produce noxious gases such as NO.sub.x or SO.sub.x, the removal of which necessitates taking additional measures to ensure that any gas ultimately released to the atmosphere is non-polluting. In addition, such high temperatures may cause the volatilization of radionuclides from the radioactive waste and vaporization of some of the constituents of the waste material. In the nuclear industry various organic amine chelating agents are utilized for cleaning the interior surfaces of the primary coolant loop of the reactor, a typical chelating agent being ethylenediaminetetraacetic acid (EDTA). Such chelating agents are used extensively for cleaning the interior surfaces of the primary coolant loop since they have an affinity for a variety of metal ions. In use, the chelating agent is used in an aqueous medium. Since the acid form of the chelating agent is substantially immiscible in water, it is common practice to add a material to increase its solubility. Typically, the material will be a sodium salt of the chelating agent. After use, the aqueous medium will also contain radioactive isotopes of various metals such as cobalt, manganese, cesium, iron etc. Heretofore there has been no truly effective way of treating such an aqueous medium. More specifically, the chelating agent contains both a source of oxygen and a source of fuel and has a relatively low decomposition temperature. Thus, treatment at any elevated temperature would result in decomposition and combustion of the chelating agent. Conversely, if treated at a lower temperature to evaporate water and reduce the volume, the resulting residue has a sticky consistency and is difficult to handle or transport. The reason is not known with certainty, but is surmised that perhaps the combination of the chelating agent, metal ions and sodium salt form a highly hydrated complex at temperatures below the decomposition temperature of the chelating agent. Typically, the aqueous medium containing the chelating agent and metal ion have a very low radioactivity and it would be acceptable to bury the solids content of the aqueous medium in drums in special, set-aside areas where ground water leakage and interaction with other radionuclides are controlled. The complex formed between the chelating agent and the metal ion, however, is water soluble. Thus, the common method for disposal of a spent aqueous medium containing a chelating agent is by solidification in cement. Obviously, this type of disposal technique will generally result in a net increase in volume. Further, the overall cost for such a disposal technique is quite high. Consequently, there is a need for a process which can be used to reduce the volume of such a radioactive waste without producing noxious off-gases or volatilizing the chelating agent or radionuclides. This need is particularly pronounced in the case of liquid low-level radioactive wastes where large volumes of wastes of relatively low radioactivity compound the problems and costs involved in their transportation and disposal. SUMMARY OF THE INVENTION In general, the present invention provides a process for reducing the volume of a low-level radioactive liquid waste containing an organic amine chelating agent by spray drying to produce a dry, flowable solid product containing the radioactive materials and chelating agent which is readily disposed of. The process broadly comprises introducing the liquid waste in the form of a finely atomized spray into a spray dryer and into intimate contact with a hot gas stream. A key aspect of the present invention is the use of a hot gas stream having a temperature in excess of the decomposition temperature of the chelating agent and controlling the proportions of the hot gas stream and liquid waste such that in a time of less than about six seconds water is rapidly evaporated from the liquid waste and the hot gas stream is cooled to a temperature below the decomposition temperature of the chelating agent. By so doing, it is possible to produce a dry, flowable powder product including the radioactive constituents of the waste and the chelating agent. There also is produced a gaseous product comprising water vapor and which is substantially free of volatile radioactive constituents from the waste. The gaseous product, after suitable purification to remove particulates, is sufficiently non-polluting to be released to the atmosphere. The powder product, which is substantially reduced in volume compared to the volume of the initial waste, is readily disposed of by conventional means such as storage or burial or incorporation into a solid matrix such as a glass, ceramic, polymeric or concrete matrix prior to storage or burial. DESCRIPTION OF THE PREFERRED EMBODIMENTS The process of the present invention accomplishes volume reduction of a low-level radioactive liquid waste which contains free water and an organic amine chelating agent by contacting such waste in the form of a finely atomized spray with a hot gas to vaporize the water from the waste. The present invention is applicable to a wide variety of organic amine chelating agents. It is particularly applicable to those more difficult to treat chelating agents such as the various organic amine acid compounds. Examples of such compounds are Ethylenediaminetetraacetic acid (EDTA), Diethylenetriaminepentaacetic acid (DTPA), Nitrilotriacetic acid (NTA) and N-Hydroxyethylethylenediaminetriacetic acid (HEDTA). Heretofore it was not believed possible that an aqueous medium containing such chelating agents in complex with metal ions could be readily dried in a short time to produce a flowable powder product. More particularly, at temperatures in excess of their decomposition temperature, even in an inert atmosphere, the compounds would decompose producing a combustible, potentially explosive, gaseous mixture. At temperatures below their decomposition temperature, after evaporation of the free water in a short residence time dryer, there is left a sticky residue which is not amenable to further processing which would require it to be passed through a conduit, pump, valve, or the like. An essential aspect of the present invention is that the hot gas and liquid waste containing the chelating agent be rapidly and intimately mixed to produce the desired powder product and cool the gas to a temperature below the decomposition temperature of the chelating agent in a time of from about 1 to 6 seconds. Thus, a spray dryer is uniquely suited for the practice of the present invention. A particularly preferred apparatus in which to carry out the process of this invention is a heated gas spray dryer in which the hot gas is produced by burning a suitable gaseous, liquid or solid fuel with an oxygen-containing gas such as air, oxygen-enriched air or oxygen in a suitable burner. The resulting hot gas is then introduced into the spray dryer at a controlled rate to provide the desired temperature in the spray dryer. Any combustible gas, such as natural gas or propane; liquid, such as fuel oil or kerosene; or solid fuel, such as coal or coke, may be used in such a burner. Fuel oil is preferred as the fuel because of its lower cost and convenience. Alternatively, the hot gas may be produced by passing air or any other gas into contact with an electrical resistance heater or in indirect contact with some heating medium. Further, in some instances it may be advantageous to use an inert gas such as CO.sub.2, N.sub.2 and the like. The initial temperature of the hot gas stream introduced into the spray dryer is a critical aspect of the present invention. Specifically, it is essential that the temperature be above the decomposition temperature of the chelating agent. If the temperature is not in excess of the decomposition temperature of the chelating agent then rather than obtaining the desired powder product there will be formed a sticky residue which will deposit on the walls of the spray dryer and the outlet ducting. Conversely of course, the temperature must not be so high that it cannot be rapidly reduced in less than about six seconds to a temperature below the decomposition temperature of the chelating agent. Thus the temperature will generally be within the range of from about 250.degree. to 400.degree. C. Particularly good results are obtained by operating with an inlet hot gas temperature of from about 300.degree. to 330.degree. C. and cooling the gas to a temperature below the decomposition temperature of the chelating agent in a time of from about 1.5 to 3 seconds. The chelating agent-metal ion complex is recovered as a dry, dense flowable powder. The powder product is well suited for situations where the waste material will ultimately be solidified in, for example, concrete or storage without solidification. In accordance with the invention, it is essential that the temperature of the hot gas be rapidly reduced to a temperature less than the decomposition temperature of the chelating agent. For convenience, the temperature is measured at the outlet of the spray dryer and should be within the range of from about 150.degree. to 200.degree. C. and preferably within the range of from about 165.degree. to 190.degree. C. In accordance with a preferred embodiment wherein the dry powder product is entrained in the gas stream and subsequently passed to a gas-solid separator such as a fabric filter, the temperature is further reduced to permit the use of conventional materials in the fabric filter. This preferably is accomplished by the introduction of dilution air at the exit of the spray dryer. In addition, since it is known that some chelating agents such as EDTA will begin to decarboxylate at temperatures as low as 150.degree. C., this has the further advantage of eliminating any possibility of such decarboxylation occurring downstream of the spray dryer. This result obviously should be avoided when it is desired to recover the chelate and metal ion as a complex. Typically the temperature of the effluent mixture of gas and product powder will be reduced to less than about 90.degree. C. Since an essential feature of the invention involves the rapid cooling of the hot gas stream, the hot gas must be intimately contacted with a finely atomized spray of the low level radioactive, liquid waste to be treated. A spray dryer is uniquely suited for this purpose. The liquid waste is introduced into the spray dryer through a spray nozzle, atomizing disc, or other distribution means. The selection of the appropriate distribution means for any given liquid waste is well within the skill of those versed in the art of spray drying. Spray drying of any of the above or any other low-level radioactive wastes, such as sludges, results in the production of a dry, flowable solid which contains the radioactive contaminants and a non-radioactive gas which, after filtering, can be released to the atmosphere as a non-polluting gas. The process of the present invention has many advantages. The waste to be processed requires no pretreatment, such as pH adjustment, in order to be dried. The spray drying process described above is not composition dependent and can handle virtually any feed material that will produce a dry product. The process may be carried out in an oxidizing atmosphere by utilizing an excess of an oxygen-containing gas; the solids produced are not decomposed or burned. This result is achieved by introducing the hot gas into the spray dryer at a temperature which is initially above the decomposition temperature of the chelating agent, and rapidly cooling the gas to a temperature which is still sufficiently high to assure that the material processed leaves the spray dryer in the form of a uniformly dry product. At the low temperatures of operation of the spray dryer in the process of the present invention, partial oxidation of the waste is avoided. Thus nitrogen-containing chelating agents are completely dried without releasing NO.sub.x which would be formed by decomposition and oxidation of the agents. Volatile fission products such as compounds of cesium or iodine in the liquid waste are contained in the solid product and not volatilized in the off-gases of the process. The solid product of the process of this invention is a dry, flowable powder which is readily transported to disposal in drums, immobilized in a monolith in a solidification system, or compressed in drums using equipment which is similar to conventional equipment used to compress solid radioactive wastes. These advantages are unique to the process of the present invention and provide an alternative to volume reduction processes currently in use for liquid wastes containing chelating agents, such as solidification of the liquid in cement with or without prior partial evaporation of the liquid.
	abstract	The invention presents a computerized method for tracking equipment repair that begins by receiving an equipment identification of an item of equipment to be repaired from a user through a graphic user interface. The invention provides the user with a list of common problems for that item of equipment (and similar equipment) and a component hierarchy for the item of equipment. The invention allows the user to browse through multiple levels of the component hierarchy and select a major component, a minor component, or a subcomponent from the component hierarchy. The invention receives diagnosis input from the user optionally selecting one of the problems and/or a component from the component hierarchy and, in response, provides the user with detailed information regarding the problem or component selected by the user. Such detailed information comprises, for each direct subcomponent of the selected component (highest level if none selected), the number of failures, the probability of failure, the mean time between failures, the occurrence of the most recent failure for each component and the next expected failure, etc. Successful prior repairs for the same problem/component are presented including tool, date, time, technician, components involved and action taken. These successful repairs are linked to textual comments regarding the repair. Comments made for what is later determined to be an ineffectual repair are linked to the subsequent successful repair information.
048662810	summary	The invention relates to an irradiation plant comprising an irradiation chamber and a conveyor system for conveyor units bearing articles for irradiation and moved past a radiation source the system comprising an even number of irradiation tracks disposed symmetrically to the radiation source and extending between a first transverse track and a second track each connecting the irradiation tracks at a respective end, and one transverse track being connected to the entry section and the exit section of the conveyor system to and from the irradiation chamber. CH-PSS 536 544 and 537 076 disclose an irradiation plant in which articles are irradiated with .gamma.-rays, e.g. for disinfection or for changing their physical properties. In the known plant, the conveyor units are moved around a radiating wall in batches, during a time preset for each batch and along the same preset path for all conveyor units. If the quantities of articles receiving the same dose of radiation are relatively large, the known device serves its purpose satisfactorily. In this case, since the change from one to another irradiation program does not have to occur too frequently, the operating costs are not excessively burdened by the change-over costs. It has been found in practice, however, that in many irradiation operations, e.g. in medicine or in the food industry or the treatment of plastics and glass, the article for irradiation often comes in small quantities, so that the irradiation plant can only be inefficiently used and loaded. In cases where the amounts of irradiated articles are small, the change-over times, as a proportion of the operating times, have a very adverse economic effect on the irradiation plant. A pamphlet dated June 1983 of Messrs. Atomic Energy of Canada Limited, PO Box 13500, Kanata, Ontaria, Canada discloses a .gamma.-irradiation plant with a suspended rail track on which the conveyor carriages carrying articles for irradiation are moved batchwise and manually to and from the irradiation chamber. The conveyor carriages and articles are all moved individually through the labyrinth along a single fixed path. A plant of this kind can be dangerous for the operator and requires expensive safety equipment. It is very complicated and uneconomic for the irradiation chamber to be simultaneously filled with articles which have to be irradiated differently. Although the residence time in the radiation area can be controlled by a dosimeter, there is a serious risk in this manually-operated plant that articles will be given the wrong doses of radiation. An object of the invention is to devise an irradiation plant in which relatively small quantities of differently-irradiated articles can be simultaneously irradiated, so that the irradiation chamber can be efficiently occupied and used even in these cases and the change-over work and time is reduced to an economically acceptable minimum. The irradiation plant must also provide high security for the operating staff. The plant must not need to be entered by staff during operation. This problem according to the invention is solved by an irradiation plant characterised in that each transverse track comprises a shift device for loading, unloading and changing over the conveyor units on the irradiation tracks and the conveyor units have control elements for presetting the path of the conveyor unit in the conveyor system and acting on sensors of a conveyor-system control means, and the control means is constructed so that whenever a conveyor unit is loaded on to or unloaded from an irradiation track, a shift device is disposed on each side of the conveyor unit, and at the end of each loading or unloading process, not more than one of the two facing shift devices is loaded with a conveyor unit. The dependent claims relate to advantageous further features of the irradiation plant. The path travelled in the irradiation chamber can be individually preset for each conveyor unit, thus enabling a number of differently-irradiated articles to travel through the irradiation chamber along various tracks depending on the preset irradiation requirement. During the loading and unloading process, only one of the two facing shift devices on the transverse tracks is occupied at one time by a conveyor unit, thus ensuring that the conveyor units run unimpeded along their individual paths through the irradiation chamber.
	summary
	abstract	A reactor internals component having a sensor insert for monitoring reactor internals components, supported within the reactor internals. The sensor insert has a hollow internal cavity that houses a number of compartmentalized, self-contained, passive environmental sensors. The sensor inserts are provided with a tracking code that can be employed to identify their location and orientation.
	summary
	claims	1. A method of determining a catalytic effect of a noble metal within a reactor of a nuclear plant, the method comprising:injecting the noble metal into a reactor water side stream, the reactor water side stream being representative of water from within the reactor, the reactor water side stream exiting the reactor and flowing to an electrochemical corrosion potential (ECP) sensor;increasing an electrochemical corrosion potential of the reactor water side stream by injecting demineralized water into the reactor water side stream to produce an oxygenated stream with an increased oxygen concentration; anddetermining the catalytic effect of the noble metal deposited within the reactor by performing a plurality of electrochemical corrosion potential (ECP) measurements on the oxygenated stream and assessing a subsequent decrease in the electrochemical corrosion potential. 2. The method of claim 1, wherein the injecting demineralized water step includes adding demineralized water with a known oxygen concentration of at least 20 times more oxygen than the reactor water side stream. 3. The method of claim 1, wherein the injecting demineralized water step includes adding demineralized water to the reactor water side stream, the reactor water side stream having less than 100 ppb oxygen. 4. The method of claim 1, wherein the injecting demineralized water step includes adding demineralized water with at least 2000 ppb oxygen to the reactor water side stream. 5. The method of claim 1, wherein the injecting demineralized water step includes adjusting a flow rate of the demineralized water such that a temperature of the oxygenated stream is at least 400° F. after injecting the demineralized water. 6. The method of claim 1, wherein the injecting demineralized water step includes adjusting a flow rate of the demineralized water such that a hydrogen-to-oxygen molar ratio in the oxygenated stream is greater than 2 after injecting the demineralized water. 7. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water at a point upstream from a clean-up system. 8. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water at a point upstream from a recirculation system. 9. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water at a point upstream from a catalytic mitigation monitoring system (MMS). 10. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water into a pipe carrying the reactor water side stream, the pipe being connected to the electrochemical corrosion potential (ECP) sensor, the adding being performed at a point upstream from the electrochemical corrosion potential (ECP) sensor, the point being at a distance of at least 10 times a diameter of the pipe. 11. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water at a flow rate that is 10% or less of a flow rate of the reactor water side stream. 12. The method of claim 1,wherein the injecting demineralized water step includes adding the demineralized water before the injection of the noble metal and while the electrochemical corrosion potential (ECP) is being measured so as to determine the catalytic effect of the noble metal. 13. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water during an injection of the noble metal and while the electrochemical corrosion potential (ECP) is being measured so as to determine the catalytic effect of the noble metal. 14. The method of claim 1, wherein the injecting demineralized water step includes adding the demineralized water after an injection of the noble metal and while the electrochemical corrosion potential (ECP) is being measured so as to determine the catalytic effect of the noble metal. 15. The method of claim 1, wherein the injecting demineralized water step includes adding demineralized water in liquid form. 16. The method of claim 1, wherein the injecting demineralized water step includes adding demineralized water that has been produced on site at the nuclear plant. 17. The method of claim 1, wherein the injecting demineralized water step includes pumping the demineralized water into the reactor water side stream with a positive displacement pump. 18. A method of determining a catalytic effect of a noble metal deposited within a reactor system, the method comprising:injecting demineralized water into a reactor water side stream to produce an oxygenated stream with an increased oxygen concentration and an increased electrochemical corrosion potential such that a hydrogen-to-oxygen molar ratio of the oxygenated stream is less than infinity the reactor water side stream being representative of water from within a reactor of the reactor system, the reactor water side stream exiting the reactor and flowing to an electrochemical corrosion potential (ECP) sensor; anddetermining the catalytic effect of the noble metal deposited within the reactor system by performing a plurality of electrochemical corrosion potential (ECP) measurements on the oxygenated stream and assessing a subsequent decrease in the electrochemical corrosion potential. 19. The method of claim 1, wherein the injecting demineralized water step includes the demineralized water having a resistivity of at least 0.1 MΩ·cm and a conductivity of at most 1 μS·cm−1. 20. The method of claim 1, wherein the injecting demineralized water step includes the demineralized water having a resistivity of at least 0.1 MΩ·cm and a conductivity of at most 1 μS·cm−1.
050646036	claims	1. A hydroball string sensing system for a nuclear reactor having a core containing a fluid at a fluid pressure, said system comprising: a tube connectable to the nuclear reactor so that the fluid can flow within said tube at a fluid pressure that is substantially the same as the fluid pressure of the nuclear reactor core; a hydroball string including-- first sensor means, positioned outside a first segment of said tube, for sensing one of said objects being positioned within said first segment, and for providing a sensing signal responsive to said sensing of said first sensing means. second sensor means, positioned outside a second segment of said tube being spaced a given distance along said tube from said first segment, for sensing one of said objects being positioned within said second segment, and for providing a sensing signal responsive to said sensing of said second sensing means. timing means for determining an amount of time between said first sensor means sensing said one of said objects within said first segment and said first sensor means sensing another one of said objects within said first segment. means for determining a velocity of said objects based upon said specified spacing and said determined amount of time; and means for adjusting the rate of the fluid flow so as to make said determined velocity substantially equal to a desired velocity. first ultrasonic transducer means for transmitting ultrasound into said tube; and second ultrasonic transducer means for receiving ultrasound having passed through said tube and for providing said sensing signal in accordance with said received ultrasound. a magnet having a magnetic field thereabout an being positioned adjacent said tube; pole pieces positioned to guide a portion of the magnetic field; and magnetic sensor means, positioned between two of said pole pieces, for sensing said portion of said magnetic field guided by said pole pieces and for providing said sensing signal in accordance with said sensed magnetic field. a magnet having a magnetic field thereabout and being positioned adjacent said tube; a pole piece positioned to guide at least a portion of the magnetic field; and coil means positioned about said pole piece for sensing the magnetic field guided by said pole piece and for providing said sensing signal in accordance with said sensed magnetic field. RF pulse generating means for driving said first ultrasonic transducer means of each of said first and second sensor means in accordance with a clock signal; a clock circuit connected to provide said clock signal; and signal processing means for amplifying and filtering said sensing signal and for providing said amplified and filtered sensing signal as an output. gate means, operatively connected between said timing means and said signal processing means, for providing said amplified and filtered sensing signal in accordance with said clock signal. detector means for detecting radiation emitted from one of said objects during a time while said one of said objects is between said first and second sensor means. detector means for detecting radiation emitted from one of said objects during a time while said one of said objects is between said first and second sensor means. detector means for detecting radiation emitted from one of said objects during a time while said one of said objects is between said first and second sensor means. a tube connectable to the nuclear reactor so that the fluid can flow within said tube at a fluid pressure that is substantially the same as the fluid pressure of the nuclear reactor core; a hydroball string including-- second ultrasonic transducer sensor means, positioned outside a second segment of said tube being spaced a given distance along said tube from said first segment, for sensing one of said objects being positioned within said second segment, and for providing a sensing signal responsive to said sensing of said second sensor means; timing means for determining an amount of time between said first sensor means sensing said one of said objects within said first segment and said first sensor means sensing another one of said objects within said first segment; means for determining a velocity of said objects based upon said specified spacing and said determined amount of time; means for adjusting the rate of the fluid flow so as to make said determined velocity substantially equal to a desired velocity; and detector means, positioned outside of said tube and between said first and second segments for counting gamma rays emitted from one of said objects while said one of said objects is between said first and second sensor means. 2. A system according to claim 1, further comprising: 3. A system according to claim 2, further comprising: 4. A system according to claim 3, wherein the fluid within said tube flows at a rate, and said system further comprises: 5. A system according to claim 4, wherein each of said first and second sensor means comprises: 6. A system according to claim 4, wherein each of said first and second sensor means comprises: 7. A system according to claim 4, wherein each of said first and second sensor means comprises: 8. A system according to claim 5, further comprising: 9. A system according to claim 8, wherein each of said first and second sensor means further comprises: 10. A system according to claim 9, wherein said hydroballs comprise stainless steel. 11. A system according to claim 10, wherein said bullet members comprise stainless steel. 12. A system according to claim 5, further comprising: 13. A system according to claim 12, wherein said detector means includes means for counting gamma rays emitted from said one of said objects. 14. A system according to claim 6, further comprising: 15. A system according to claim 7, further comprising: 16. A system according to claim 6, wherein said hydroballs comprise ferritic stainless steel. 17. A system according to claim 16, wherein said bullet members comprise ferritic stainless steel. 18. A system according to claim 7, wherein said hydroballs comprise ferritic stainless steel. 19. A system according to claim 18, wherein said bullet members comprise ferritic stainless steel. 20. A hydroball string sensing system for a nuclear reactor having a core containing a fluid at a fluid pressure, comprising:
	claims	1. An aerosol generating and mixing system operating at a high temperature and a high pressure, comprising:an aerosol generating device; andan aerosol mixing device,wherein the aerosol generating device includes a pre-mixing tank and a mixing tank, and the mixing tank and the pre-mixing tank include a wing configured to rotate about a central shaft of the tank so as to agitate an inside aerosol, and an agitating motor configured to rotate the wing, and a filling nozzle of the mixing tank and the pre-mixing tank is configured to inject any of an aerosol aqueous solution and an aerosol particle, and the aerosol aqueous solution discharged from the pre-mixing tank is injected in the mixing tank, and a first air injector is provided to inject any of a compressed air and a nitrogen gas into the pre-mixing tank and the mixing tank so as to pressurize the tank, and an aerosol aqueous solution is discharged with the aid of a gear pump while preventing the aerosol from hardening in the aerosol aqueous solution by means of agitation inside the mixing tank and the pre-mixing tank, and a predetermined quantity of the aerosol aqueous solution can flow using a mass flow meter, and an aerosol mixture can be produced by combining the aerosol aqueous solution passing through the mass flow meter and the air of a second air injector which is able to inject any of the compressed air and the nitrogen gas which are inputted through a separate line and by injecting the produced aerosol mixture into a second tank, thus generating an aerosol. 2. The system of claim 1, wherein the aerosol mixture is injected through a binary fluid nozzle configured to inject air and an aerosol together when injecting into the second tank, and the aerosol mixture discharged from the binary fluid nozzle is able to form a previously set pattern inside the second tank. 3. The system of claim 2, wherein the aerosol mixture discharged from the mixing tank passes through a gear pump and is injected in a high pressure state and is heated while it passes through a heater and is injected in the second tank, and the air discharged from a second air injector is heated while it passes through the heater, so the air is injected in a heated state. 4. The system of claim 3, wherein a mixing ring is provided inside the second tank and where a fine spray sprayed through the binary fluid nozzle is formed, and one or more than one of vapor, a compressed air, a nitrogen gas and a mixture thereof is discharged through a transfer gas spray hole which is formed at an outer circumferential surface as it is inputted from the top of the mixing ring in a donut shape. 5. The system of claim 4, wherein a mixing ring block part is provided at a portion symmetrically matching with where the mixing gas is inputted, so as to maintain a uniform transfer gas flow speed distribution at the upper and lower parts of the mixing ring.
	claims	1. A system for collecting 3He, the system comprising:a) a nuclear reactor having a vessel containing heavy water and having a cover gas head space containing a cover gas above the heavy water and a gas outlet in communication with the cover gas head space, whereby operation of the nuclear reactor results in thermal neutron activation of deuterium in the heavy water to produce tritium (3H) and at least some of the tritium undergoes β− decay to produce 3He gas that mixes with the cover gas;b) a gas extraction passage fluidly connected to the gas outlet of the vessel to extract a gas outlet stream through the gas outlet, the gas outlet stream comprising the cover gas and the 3He gas mixed with the cover gas;c) a 3He separation apparatus fluidly connected to the gas extraction passage downstream gas outlet and operable to receive the gas outlet stream and separate the 3He gas from the cover gas wherein the 3He separation apparatus comprises at least one of a thermal diffusion apparatus, a fractional diffusion apparatus, a heat flush apparatus, a superleak apparatus and a differential absorption apparatus. 2. The system of claim 1, further comprising a gas inlet provided in the vessel and in communication with the cover gas head space, and a cover gas supply passage coupled to the gas inlet of the vessel to supply the cover gas to the cover gas head space. 3. The system of claim 2, wherein the 3He separation apparatus comprises a 3He outlet to output a separated 3He gas stream and a separate treated cover gas outlet to output a treated cover gas stream. 4. The system of claim 2, wherein the treated cover gas outlet of the 3He separation apparatus is fluidly connected to the cover gas supply passage to re-introduce at least a portion of the treated cover gas stream into the cover gas head space. 5. The system of claim 1, wherein the gas outlet stream is extractable as a generally continuous gas stream while the nuclear reactor is in operation. 6. The system of claim 1, wherein the cover gas provided above the heavy water consists essentially of 4He. 7. A moderator cover gas system for use with a nuclear reactor having a vessel containing heavy water, the cover gas system comprising:a) a cover gas supply passage having a gas outlet connectable to a gas inlet on the vessel to supply a cover gas into the vessel,b) a gas extraction passage having a gas inlet connectable to a gas outlet on the vessel to extract an outlet gas stream from within the vessel, the outlet gas stream comprising a mixture of at least the cover gas and 3He gas;c) a gas separation apparatus connected to the cover gas flow passage downstream from the gas outlet on the vessel and operable to separate the 3He gas from the outlet gas stream, wherein the 3He separation apparatus comprises at least one of a thermal diffusion apparatus, a fractional diffusion apparatus, a heat flush apparatus, a superleak apparatus and a differential absorption apparatus. 8. The system of claim 7, further comprising a fresh cover gas source fluidly connected to the cover gas supply passage to introduce cover gas consisting essentially of 4He into the interior of the vessel. 9. The system of claim 7, wherein the gas separation apparatus comprises a first outlet to output the separated 3He gas and a second outlet to output a treated cover gas stream. 10. The system of claim 9, wherein the second outlet is fluidly connected to the cover gas supply passage to feed at least a portion of the treated cover gas stream into the cover gas supply passage.
	abstract	An apparatus for performing an inspection on the beams of the top guide of a BWR includes a housing, an alignment assembly, and an inspection system. The housing is receivable atop the upper edges of a first pair of beams adjacent a receptacle of the top guide. The reception of the housing atop the upper edges of the first pair of beams is facilitated by the alignment assembly which includes a plurality of legs that are simultaneously moved between a retracted position wherein one or more of the legs is disengaged from the beams within the receptacle and an extended position wherein all of the legs are engaged with the beams of the top guide within the receptacle. The inspection system includes a pair of inspection elements that are translated above a second pair of beams that are adjacent the receptacle and that do not have the housing received thereon.
	description	This application claims the benefit of U.S. Provisional Application No. 61/586,534, filed Jan. 13, 2012, which is hereby fully incorporated by reference. The present invention relates to radiation shielding from Gamma and X-Ray sources (Gamma). It finds particular application in conjunction with using different sizes of material to obtain shielding at reduced thickness and/or total weight per unit volume. It will be appreciated, however, that the invention is also amenable to other applications and will be described with particular reference thereto. Current regulatory and environmental concerns and events worldwide have created an emergent need for a more effective radiation shield that may be deployed in previously unrecognized fashions. Man made and natural disasters are at an all time high. From earthquakes and blasts, to dirty bombs, and nuclear weapons (radiation), it is evident that there is a need for protective materials that cover a multitude of varying radiological threat assessments and environments. Many products have been invented with a common goal of protecting human life and infrastructure. It should be further stated that environmental impact is a consideration in new infrastructure development and product manufacturing. Products such as concrete, lead, and steel coatings have been used in various applications to protect human life and infrastructure from ionizing radiation. The use of lead and/or concrete, for example, for protection against multi-spectral-radiations has resulted in ecological debates regarding the creation and disposal of lead based products. The present invention provides a new and improved apparatus and method which addresses the above-referenced problems. In one embodiment, a radiation shielding panel includes a tungsten powder and a polyurea material. The tungsten powder includes tungsten particles having three different specific diameters. The tungsten powder is mixed and dispersed into the polyurea material. The mixture of the polyurea material and the tungsten powder shields radiation greater than about 6 MeV. With reference to FIGS. 1 and 2, a simplified component diagram of an exemplary wall panel 10 is illustrated in accordance with one embodiment of the present invention. In the illustrated embodiment, the panel 10 includes one (1) layer. The layer is a combination of an “A” component 12 (e.g., “A″ side”) and a “B” component 14 (e.g., “B″ side”). In one embodiment, the A component 12 includes an accelerant, and the B component 14 represents a main body of the panel 10 and includes a polyurea mixture (polyurea material). The panel 10 is made by combining the A component 12 and the B component 14 with a particle mixture of tungsten. In one embodiment, the mixture of tungsten particles is a tungsten powder. The A component 12, the B component 14, and the particle mixture of tungsten powder are then poured into a form and allowed to cure. In one embodiment, the main body 14 of the panel 10 is HM-VK™ ultra high strength handmix polyurea elastomer. With reference to FIGS. 1-3, during a first phase, in a step 50, the B component 14 polyurea mixture of a predetermined amount (e.g., about 20 lbs. to about 25 lbs. of the Tungsten powder and about 75 ounces to about 83 ounces of the “B” component (e.g., HM-VK™)) is placed into a mixing apparatus and stirred, by itself, for about 1 minute to about 4 minutes. In a step 52, the mixture of tungsten powder is slowly added to the polyurea mixture. The tungsten powder and the polyurea mixture are stirred together for an additional time (e.g., between about 1 minute and about 4 minutes), during which time the tungsten powder is dispersed in the polyurea mixture and the polyurea mixture and the tungsten powder mixture begin to slightly thicken into the B component 14 material. When the polyurea mixture and the tungsten powder mixture have sufficiently thickened into the B component 14 material, (e.g., after the additional stirring time of between about 1 minute and about 4 minutes), in a step 54 the A component 12 material (e.g., accelerant) is added. In one embodiment, about 20 ounces to about 25 ounces of the “A” component HM-VK™ accelerant is added in the step 54. Adding the A component 12 material causes a long chain polymer mixture to rapidly setup. The exemplary amounts of the A component 12 and the B component 14 discussed in the present paragraph have been found to result in a panel 10 of about 31.875″ by about 31.875″ by about 0.125″ thick. After about 4 minutes of stirring the combined A component 12 (e.g., the accelerant) and the B component 14 (e.g., the main panel body including the polyurea mixture and the tungsten powder mixture) in a step 56, the combined material is poured into a form and pressurized in a step 60. In one embodiment, the combined material is pressurized to about 6,000 psi for between about 4 hours and 4½ hours. After about the first seven (7) minutes of the about 4 (four) hour pressurization, the combined A component 12 and B component 14 material enters a first transitional phase (e.g., the t-phase). Then, over the remainder of the 4 hour press time (e.g., after reaching the first t-phase), the panel material attains the physical characteristics of heavy rubber. The material state during and after mixing represents numerous applications in manufacturing and emergency radiological mitigation. It is noted that the A component 12 (e.g., the accelerant) vaporizes during the four (4) hour pressurization. Companies such as, but not limited to, ArmorThane, LineX, and Specialty Products Inc. (SPI) have produced Urethane products/Polyurea products. As stated by the Polyurea Development Association, the advantages and benefits of Polyurea include: no volatile organic compounds (VOC's) and little to no odor, some systems are USDA and potable approved; weather tolerant (cures at about 25° F. to greater than about 300° F., even in high humidity); excellent resistance to thermal shock; flexible; bridges cracks; waterproof, seamless and resilient; unlimited mil thickness in one application; spray, hand mix, and caulk grade materials; excellent bond strengths to properly prepared substrates; resistant to various solvents, caustics, and mild acids; and low permeability, excellent sustainability. Material ListPolyurea B componentabout 9% to about 30% by volumeTungsten Powderabout 70% to about 93% by volumeA component about 25% to about 30% of B component by volume The percentages listed above are merely examples and, furthermore, may vary slightly for many reasons to include, but are not limited to, ambient temperature, variations in the materials used, water content, radiation shielding requirements etc. In one embodiment, the tungsten powder includes a plurality of different sized particles. In one embodiment, the tungsten powder includes three (3) differently sized particles. For example, the three (3) differently sized particles have diameters of about 90.0 microns, about 9.0 microns, and about 0.9 microns. In one example, the tungsten powder includes, by volume, about 80% of tungsten particles having a diameter of about 90.0 microns, about 15% of tungsten particles having a diameter of about 9.0 microns, and about 5% of tungsten particles having a diameter of about 0.9 microns. With reference to FIG. 4, the largest sized particles 20a (e.g., having a diameter of about 90.0 microns) establishes a base layer and then smaller sizes 20b (e.g., having a diameter of about 9.0 microns) and 20c (e.g., having a diameter of about 0.9 microns) fill in the gaps between the largest size so that final product has less weight, while providing significant shielding to radiation greater than about 6 million-electron volts (MeV), when compared to conventional materials (e.g., lead, steel, and concrete). Gamma radiation is a hazard from multiple sources including, but not limited to, nuclear weapons, mixed nuclear waste, and others. In one embodiment of the present invention, gamma radiation exposure is reduced by incorporating a shielding material, which is in the form of the panel 10 described above, that absorbs a wide variety of radiation from various gamma emitting sources. Various high atomic weight materials (e.g., concrete, steel, lead, tungsten, depleted uranium, etc.) have been used for gamma radiation shielding. In one embodiment, it is contemplated that the proprietary gamma radiation shielding materials and gamma radiation absorbers are incorporated into the polyurea layer to reduce or minimize gamma radiation penetration in the panel 10. It is contemplated that the panel 10 as shown in FIGS. 1 and 2 is about 30″ wide (see “W” in FIG. 1) by about 30″ high (see “H” in FIG. 1) with a thickness (e.g., depth) of about ⅛″ (see “D” in FIG. 2). The panel 10 includes between about 78% and about 92% of the tungsten powder and between about 8% and about 22% of the polyurea material. A panel 10′ illustrated in FIGS. 5 and 6 is about 30″ wide (see “W” in FIG. 5) by about 30″ high (see “H” in FIG. 5) with a thickness (e.g., depth) of about ¼″ (see “D” in FIG. 6). The panel 10′ includes between about 80% and about 90% of the tungsten powder and between about 10% and about 20% of the polyurea material. Such panel measurements are in accordance with ANSI 42.46 For Determination of Imaging Performance of X-Ray and Gamma Ray Systems for Cargo and Vehicle Screening. With reference to FIGS. 1, 2, 5, and 6, it is contemplated that a panel 10 having a thickness of about ⅛″ has a density of about 5.06 g/cm3, and that a panel 10′ having a thickness of about ¼″ has a density of about 5.12 g/cm3. Testing has shown that the panel 10 having a mean thickness of about ⅛″ offers radiation shielding (e.g., gamma radiation shielding) equivalent to a steel sheet having a mean thickness of about 0.38004″. Testing has also shown that the panel 10′ having a mean thickness of about ¼″ thick offers radiation shielding (e.g., gamma radiation shielding) equivalent to a steel sheet having a thickness about 0.657907″. The panel 10 includes multiple components of polyurea and tungsten and protects against multiple ionizing radiation sources. The panel 10 may be referred to as a single layer, ionizing radiation shield panel (e.g., a multi-spectral ionizing radiation panel (MSIRP) panel), including the plurality of different sized particles. It is contemplated that the panel 10 is non-structural, allowing it to be manufactured in multiple shapes and sizes to accommodate retrofitting on existing structures (e.g., embassies, federal installations, perimeter structures, refineries, etc). A non-structural panel 10 is not part of the structure of the building and maybe used as an after-market product (or addition) to the structure. The panel 10 may be described as a monolithic-like structure. The panel 10 is also pliable. As illustrated in FIG. 7, in one embodiment, one side 10a of the panel 10 is folded to be adjacent a second side 10b of the panel 10 with a clamp 22 exerting pressure of about 600 lbs./in2 proximate to a fold 24 created in the panel 10. FIG. 8 illustrates the panel 10 of FIG. 7 after being held in the clamp 22 (see FIG. 7) for about 24 hours and then removed. Two (2) lines 26 on the panel 10 approximately illustrate the position of the fold 24. FIG. 8 illustrates the pliability of the panel 10, since the panel 10 includes no cracking or fatigue at the fold 24 even after being held in the clamp 22 under about 600 lbs./in2 for about 24 hours. It is contemplated that a properly formulated panel 10, when secured adequately to a structure or testing apparatus, will withstand at least: 1. Most levels of ionizing gamma radiation (about 100 electron volts (eV) to above about 6 MeV). 2. Shock and related stressor's. The panel 10, as described above with reference to FIGS. 1, 2, and 5-7 is designed and engineered to absorb and block ionizing radiations across the spectrum from about 10 electron Volts to about 6 MeV and to protect human life and infrastructure. The panel 10 also offers flexibility in engineering and design so that panel dimensions and thickness may be adjusted to meet various radiological environments and situations based on consumer and manufacturing demands. In addition, the panel 10, as described above, is designed for a broad range of multi-purpose uses. For example, the panel 10 materials in their liquid state, may be sprayed and applied by conventional methods for providing protection for nuclear based applications ranging from nuclear power plants, medical, radiological laboratories and other locations such as military applications and also for homeland security. Still other uses would include substitution for traditional lead shielding in commodities including x-ray machines, medical radiology suites and others. In each of these uses, the concept is to provide significant protection for employees from the effects of crime, terrorism or warfare. The potential threats from small arms, car bombs and improvised devices which might include radiation dispersal devices are the most likely security and military based use. The end result of using this material is that a wide range of protections are available from the same material (e.g., the panel 10). In one embodiment, the panel 10 shields radiation greater than about 6 MeV. Tungsten and polyurea have not been used in radiation shielding mix designs, primarily due to the inherent cost of pure ore. In the embodiments of the present invention, we have tested and categorically shown that better gamma radiation shielding can be attained using tungsten powder, in lieu of relatively expensive sheets of tungsten—a recent test we have conducted confirms this. During the curing process, the panel material in the press reaches the first t-phase relatively rapidly (e.g., in minutes). Then, as noted above, over the remainder of the 4 hour press time (e.g., after reaching the first t-phase), the panel material attains the physical characteristics of heavy rubber. Overpressure from explosives, chemical, mechanical and other events are often the part of an event with the greatest hazard. The panel 10 provides protection for people or systems (e.g., an office and/or electronic equipment) to enhance survivability. By coupling several kinds of protection, from ionizing radiation and radio frequency radiation, the panel 10 described above fulfills several requirements for security and protection simultaneously and in a new and unique way. While the present invention has been illustrated by the description of embodiments thereof, and while the embodiments have been described in considerable detail, it is not the intention of the applicants to restrict or in any way limit the scope of the appended claims to such detail. Additional advantages and modifications will readily appear to those skilled in the art. Therefore, the invention, in its broader aspects, is not limited to the specific details, the representative apparatus, and illustrative examples shown and described. Accordingly, departures may be made from such details without departing from the spirit or scope of the applicant's general inventive concept.
048037168	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS A high voltage supply 1 is shown in the drawing which feeds an x-ray tube 2, and which is connected to the mains via a line 3. The x-ray tube 2 emits an x-ray beam 4 having a central ray indicated at 10. The x-ray beam 4 during a radiograph exposure penetrates an examination subject 5, passes through a radiation measuring chamber 6, passes through a secondary radiation grid 7, and is incident on an x-ray film 8 in an x-ray film cassette 9. During an exposure, the secondary radiation grid 7 is moved perpendicularly to the central ray 10 by a motor 11, so that the lamellae of the grid 7 which are directed at the focus of the x-ray tube 2, do not form an image on the film 8. An electrical output of the radiation measuring chamber 6 is connected to an input of an integrator 13 through an amplifier 12. The output of the integrator 13 is an electrical signal corresponding to the actual value of the radiation dose, and is supplied to one input of a comparator 14. Another input of the comparator 14 is supplied with an electrical signal on a line 15 corresponding to the rated value of the radiation dose which is required for an optimum film blackening. When the actual value and the rated value of the radiation dose are the same, the comparator 14 supplies a signal to the high voltage supply 1 which de-energizes the supply and thus shuts off the x-ray tube 2. The input signal of the integrator 13 (the current i) is proportional to the dose rate D. The exposure time is fixed by this signal. This signal is supplied through an amplifier 17 to a motor control unit 18 which controls the speed of the motor 11 and thus the speed of movement of the secondary grid 7. The signal on line 15 is also supplied to the amplifier 17 and controls the amplification factor thereof. The adjustment speed of the secondary radiation grid 7 is shown in the drawing as X.sub.R. This adjustment speed X.sub.R will be higher as the anticipated exposure times becomes shorter. The rated value and the actual value of the dose are calculated according to the following equation: ##EQU1## wherein T is the exposure time, D.sub.actual is the actual value of the dose rate and t is real time. The following relationship derives for the path X.sub.R traversed by the secondary radiation grid: ##EQU2## wherein X.sub.R, as above, is the speed of the secondary radiation grid. This speed can be determined according to the following relationship: ##EQU3## The dose rate can be maintained constant during the exposure time, or may be a function of the time. According, X.sub.R is then constant or a function of time t. For adequate blurring, a path distiance X.sub.R which the secondary radiation grid 7 must cover during an x-ray exposure is fixed. In order to avoid differences in optical density due to the secondary radiation grid 7 being off-centered, one half of the path distance X.sub.R is present on each side of the central ray 10. Instead of controlling the speed of the secondary radiation grid as discussed above, a control circuit for this speed wherein the rated value is fixed dependent on the anticipated exposure tme may also be provided. Although modifications and changes may be suggested by those skilled in the art it is the intention of the inventors to embody within the patent warranted hereon all changes and modifications as reasonably and properly come within the scope of their contribution to the art.
043572979	claims	1. In a nuclear reactor including a primary tank having a vertically extending annular side wall and containing therein a top hot pool and a bottom cold pool of coolant separated from one another by a horizontally extending barrier, an arrangement for thermally insulating the inner circumferential surface of said tank side walls, said arrangement comprising: (a) a cylindrical metal liner located concentrically within said tank and spaced inwardly of said inner circumferential surface in confronting relationship therewith whereby to define an axially extending annular space therebetween, said space being filled with an inert gas and having an opened bottom end in fluid communication with said cold pool whereby coolant leaking into said space from either of said pools will drain to said bottom open end; (b) means for hermetically sealing the top end of said space; (c) a plurality of vertically extending, radially spaced apart, reflective metal plates having arcuate cross-sections located within and together extending the entire circumferential length of said annular space for reducing radiative heat transfer radially across said space, thereby thermally insulating said tank side wall from the temperatures generated by the reactor, each of said plates having a plurality of indentations in its surface for providing stand-off spacing between adjacent plates and thereby reducing convective heat transfer radially across said space and for further facilitating drainage of any liquid coolant which appears between said plates; and (d) means for supporting said plates in position within said space. (a) a cylindrical metal liner located concentrically within said tank and spaced inwardly of said inner circumferential surface in confronting relationship therewith whereby to define an axially extending annular space therebetween, said space being filled with inert gas and having an opened bottom end in fluid communication with said cold pool; (b) means for hermetically sealing the top end of said space; (c) a plurality of vertically extending, radially spaced apart, reflective metal plates having arcuate cross-sections located within and extending the length of said annular space for thermally insulating said tank side wall from the temperatures generated by the reactor, each of said plates having a plurality of indentations in its surface for providing stand-off spacing between adjacent plates and for reducing convective heat transfer between points in said reactor and said primary tank; (d) means for supporting said plates in position within said space; (e) first tubular means extending into said space from outside said tank for injecting inert gas into said space; and (f) second tubular means extending into said space from outside thereof for monitoring the gas pressure within said space. 2. In a nuclear reactor including a primary tank having a vertically extending annular side wall and containing therein a top hot pool and a bottom cold pool of coolant separated from one another by a horizontally extending barrier, an arrangement for thermally insulating the inner circumferential surface of said tank side walls, said arrangement comprising: 3. An arrangement as in claim 2 wherein each of said first and second tubular means includes at least one tube extending into said space from outside thereof and wherein one of said tubes is concentrically positioned around the other, said concentrically positioned tubes extending through a cooperating opening in said lining. 4. An arrangement as in claim 2 including a plurality of additional vertically extending, radially spaced apart, reflective metal plates shorter axially than said first-mentioned plates and liner extension means positioned on the innermost surface of said liner and integrally formed with a top annular section of said liner for supporting said additional plates in position adjacent to said liner. 5. An arrangement as in claim 4 including a plurality of second additional vertically extending, radially spaced apart, metal plates, shorter axially than said first mentioned plates and second liner extension means positioned on the innermost surface of said liner and integrally formed with a bottom annular section of said liner for supporting said second additional plates in position adjacent to said liner. 6. An arrangement as in claim 2 wherein each of said reflective plates has a rigid supporting member attached along its upper margin, said members along with the reflective plates being connected with a plurality of radially positioned bracket members connected with said tank, said bracket members having T-shaped cross sections, said support members and bracket members serving as said supporting means. 7. An apparatus as in claim 2 wherein the reflective plates are vertically segmented into arcuate sections. 8. An apparatus as in claim 2 wherein the nuclear reactor is cooled by a flow of liquid metal and the reflective plates are fabricated from stainless steel.
040452824	summary	This invention is directed to a method of thermal monitoring of a nuclear reactor core and to a monitoring device for carrying out said method. It is known that, in the case of a nuclear reactor which is in operation, temperature is a physical parameter which has to be checked with care since this latter must not be permitted to depart from a predetermined value at a given monitoring point of the reactor, said parameter being a function of the materials located in the proximity of said monitoring point. Thus the coolant temperatures within the fuel assemblies must always be such that the clad temperature does not exceed a predetermined maximum value under any circumstances. Prior to the present invention, it was the customary practice to carry out thermal monitoring of a reactor core by means of the temperature rise produced within the coolant as this latter passed through the fuel assemblies by detecting successively on the one hand the temperature of the coolant at a point located upstream of the fuel assemblies and on the other hand the temperatures of the coolant at the outlet of each fuel assembly. It is worthy of note that the periodic application of this method did not permit fast detection of a possible sudden temperature rise occurring in a fuel assembly between two successive measurements carried out on said fuel assembly and that the accuracy of a method of this type was also liable to be affected when a substantial temperature variation occurred upstream of the fuel assemblies. The present invention is precisely directed to a method of thermal monitoring of a nuclear reactor core which overcomes the drawbacks mentioned in the foregoing since it has the advantage in particular of permitting continuous monitoring of each reactor fuel assembly. In more exact terms, the method of thermal monitoring of a reactor core in accordance with the invention and comprising "hot", sometimes hereinafter referred to as T.sub.1, and "cold", sometimes hereinafter referred to as T.sub.2, fuel assemblies by utilizing coolant temperatures detected at the outlets of a plurality of fuel assemblies essentially consists: IN ESTABLISHING AT EACH INSTANT ON THE ONE HAND THE MEAN CORE OUTLET TEMPERATURE BY FORMING THE HALF-SUM OF THE MEAN "HOT" (T.sub.1) and "cold" (T.sub.2) core outlet temperatures corresponding respectively to the mean value of the coolant temperatures at the outlet of at least part of the hot fuel assemblies and to the mean value of the coolant temperatures at the outlet of at least part of the cold fuel assemblies and, on the other hand, the difference between said mean "hot" (T.sub.1) and "cold" (T.sub.2) core outlet temperatures, in producing for each fuel assembly aforesaid an analog signal corresponding to the temperature difference between the coolant temperature at the outlet of said assembly and the mean core outlet temperature as initially increased or decreased by a predetermined fraction of the difference between the mean "hot" (T.sub.1) and "cold" (T.sub.2) core outlet temperatures in order to define a state of equilibrium at which said temperature difference is equal to zero, in processing said signal in order to ensure that the appropriate safety actions are initiated by overstepping of threshold values corresponding to temperature variations permitted on each side of the equilibrium zero. The method as defined in the foregoing makes it possible to carry out continuous relative monitoring of the temperatures of a reactor core without entailing the need for continuous comparison with the coolant inlet temperature since it consists in detecting at each instant any potential localized cooling defect in one of the reactor core assemblies by comparison with the cooling at the same instant of the reactor core considered as a whole. Said zero method consists in continuously comparing the coolant outlet temperature of each of the core monitoring assemblies with the mean outlet temperature of the core considered as a whole by initially establishing an analog state of equilibrium such that the difference between these two compared temperatures is reduced to zero by a fraction of the difference between the mean "hot" (T.sub.1) and "cold" (T.sub.2) core outlet temperatures, the fraction aforementioned being intended to constitute a characteristic parameter of each fuel assembly since said fraction is a function of the position of the fuel assembly considered within the reactor core. In more exact terms, the method according to the invention can be presented by way of explanation as follows: if a signal u.sub.k represents the coolant outlet temperature of the k.sup.th fuel assembly which is detected for example by a thermocouple at the outlet of said k.sup.th assembly, if a signal u.sub.c produced by the analog method from a certain number of signals u.sub.k corresponding to "hot" fuel assemblies of the reactor core represents the mean "hot" core outlet temperature, if a signal u.sub.f produced by the analog method from a certain number of signals u.sub.k corresponding to "cold" fuel assemblies of the reactor core represents the mean "cold" core outlet temperature, the method consists in following the progressive variation of the signal: ##EQU1## where u.sub.c + u.sub.f /2 provides a measurement of the mean core outlet temperature U.sub.M and c.sub.k, this characteristic of the k.sup.th fuel assembly being established at the time of an initial adjustment in order to ensure that the signal E.sub.k is of zero value. Any cooling fault which occurs in one of the fuel assemblies and modifies the signal u.sub.k to an appreciable extent results in a state of unbalance which, depending on its magnitude, automatically initiates the necessary safety actions such as a pre-alarm, an emergency shutdown or a control rod drop. It is worthy of note that, if the fuel assembly of the order K which is subjected to abnormal cooling takes part in the generation of the signals U.sub.M and u.sub.c - u.sub.f, these signals are disturbed only to a slight extent. The present invention is also directed to a device for carrying out the method in accordance with the invention which comprises in a general manner: N thermocouples which are equal in number to the number N of reactor core fuel assemblies to be monitored, N measuring channels connected individually to each of the N thermocouples and each comprising in series: at least one amplifier for the signal detected by the thermocouple, PA1 a summing amplifier at the output of the preceding amplifier and also at the output of two associated analog circuits supplied by at least a certain number of the N measuring channels and each comprising means enabling one circuit to deliver a signal corresponding to the opposite value of the mean core outlet temperature and enabling the other circuit to deliver a signal corresponding to the appropriate fraction of the difference between the mean "hot" (T.sub.1) and "cold" (T.sub.2) core outlet temperatures, PA1 at the output of said summing amplifier, devices comprising different thresholds for pre-alarm, emergency shutdown and/or control rod drop, each device aforesaid being intended to deliver command signals for appropriate orders by means of at least one regrouping OR-circuit in the event of overstepping of the corresponding threshold. Further properties and advantages of the invention will become more clearly apparent from the following description of one practical example of the method which is given by way of illustration and not in any limiting sense. In this example of application of the invention, thermal monitoring of the core of a fast reactor cooled by a liquid metal and especially sodium is carried out in a wholly satisfactory manner and offers a considerable advantage since monitoring of the "hot point" of fuel cans is of major importance in this type of reactor. The core of a nuclear reactor of this type is described in particular in the June 1973 issue of "Bulletin d'Informations Scientifiques et Techniques" No 182 published by Commissariat a l'Energie Atomique.
045089698	abstract	The invention relates to a device for holding, transporting and final storing burned-out reactor fuel elements comprising a hollow-cylindrical container that can be closed with a cover. On its sealing surface which is opposite the cover, the container is provided with projections which are of dovetail profile. The cover is cast onto the container around the dovetail projections by means of a casting mold whereby an intimate and firm connection between the container jacket and the cover can be produced. The cover can also be prefabricated with filler channels for directing metal casting material into recesses provided in the sealing surface of the cover. When the metal hardens in the recesses the cover is securely locked to the container. The container may have a shielding cover beneath the top cover.
052992462	summary	FIELD OF THE INVENTION The invention described herein relates to nuclear reactor fuel assemblies and more particularly to fuel assembly supporting structures comprising shape-memory alloys. BACKGROUND OF THE INVENTION Commercial nuclear reactors used for generating electric power include a core composed of a plurality of fuel assemblies which generate heat for electric power generation purposes. Each fuel assembly includes an array of fuel rods which are held in a spaced relationship with each other by means of spacer grids of egg-crate configuration spaced along the fuel assembly length. The fuel rods may be approximately 0.5 inch in diameter and about 12 feet long, thus requiring a number of supporting grids along their length. As discussed in U.S. Pat. No. 5,024,807 to Hatfield et al., the disclosure of which is herein incorporated by reference, metallic debris in the coolant which collects or is trapped in fuel rod spacer grids adjacent to the fuel rod cladding is believed to be responsible for a significant percentage of known fuel rod failures. Traditional fuel assembly designs were known to sustain a distribution of debris-induced failures that clearly showed the lowest spacer grid to be a very effective filter for debris. Unfortunately, the short, lower end caps on the fuel rods of such fuel assemblies ensured that the hollow cladding tubes would be adjacent to the trapped debris, and that any flow-induced motion of the debris could wear through the thin wall of the tubes and cause rod failure. In traditional fuel assemblies, the lowest spacer grid was some distance up from the bottom of the fuel rod, since, in the absence of a positive axial capture device for the rod, the grid needed to be located at an elevation where it always laterally captured a "lifted" rod. Rods could potentially lift in response to coolant flow during abnormal conditions. One choice for a debris-resistant fuel assembly design is to merely lengthen the solid end cap such that it is extends up through the bottom spacer grid. This simplistic solution, however, is not feasible because zirconium alloy bar stock used for end caps is very expensive and because void volume within the fuel rod and/or the active fuel length would be negatively affected. In response to this observation, Hatfield et al. designed a fuel assembly with a spring detent spacer grid of intersecting strips. The spring detent spacer grid allowed the grid to be moved downward, thereby reducing the solid zirconium alloy material length required. To preclude the "rod lift", the grid included a fuel rod capturing spring detent device. This device engaged a circumferential groove with tapered sides in the fuel rod end cap which created enough axial restraint to prevent or minimize rod lift under all flow conditions, but not enough restraint to significantly affect fuel rod reconstitution. In addition, integral leaves substantially symmetrically arranged on either side of the strip intersections were also added to greatly increase the likelihood that debris that passed the novel first or bottom spring detent spacer grid was too small to become trapped at a higher grid where it could damage the cladding of the active fuel region. Despite these advantages, the spring detent feature was discovered to restrict rod replacement somewhat as compared to a standard lower spacer grid without a detent feature by requiring more pulling force to be applied to the rod during reconstitution. Other spacer grids make use of conventional springs to hold fuel rods in place. For example, as discussed in U.S. Pat. No. 4,389,369 to Bryan, the disclosure of which is herein incorporated by reference, grid designs commonly include interwoven Inconel or Zircaloy straps which form multiple cells, each cell having springs on two adjacent walls and projections or dimples on each of the other two walls. The springs laterally impress resistive forces on each fuel rod in the assembly. Although this fuel assembly design performs exceptionally well in a nuclear reactor, one disadvantage inherent in the design is that the inwardly projecting springs and dimples occasionally mar or score the surface of fuel rods while they are being pulled into the fuel assembly grids. In carrying out this fuel rod loading operation, the grids are immovably held in position while a longitudinal steel rod attached to the end of a fuel rod pulls the fuel rod axially through the aligned openings, or cells, in the grids. As the rod engages the springs and dimples in the grid cells, their edges engage the exposed, relatively soft surface of the moving fuel rod and, in some cases, score its surface sufficiently deep as to cause the rod to fall outside established fuel rod surface specifications. SUMMARY OF THE INVENTION In view of the foregoing, it is readily apparent that, in prior art grid designs, the springs tend to hold the fuel rods too tenaciously during fuel rod reconstitution and/or tend to score the fuel rods during installation. One advantage of the present invention is the elimination or reduction of the above problems encountered with prior art grids. This and other advantages have been achieved by using shape-memory alloys in the construction of part or all of the nuclear fuel assembly support structures. For example, according to an embodiment of the present invention, a support structure for supporting one or a plurality of nuclear reactor components is provided. The support structure comprises a two-way shape-memory alloy having an overall transition temperature range substantially above atmospheric temperature and substantially below a temperature experienced by the shape-memory alloy during reactor operation such that the support structure assumes a first configuration for securely supporting the components at a temperature above the overall transition temperature while assuming a second configuration for loosely engaging the components below the overall transition temperature. On further study of the specification and appended claims, further advantages of this invention will become apparent to those skilled in the art.
	summary
	abstract	Process for manufacturing a nuclear component that includes i) a support containing a substrate based on a metal, the substrate being coated or not coated with an interposed layer positioned between the substrate and at least one protective layer and ii) the protective layer composed of a protective material including partially metastable chromium; the process includes a step a) of vaporizing a mother solution followed by a step b) of depositing the protective layer onto the support via a process of chemical vapor deposition of an organometallic compound by direct liquid injection (DLI-MOCVD).
046776524	abstract	There is disclosed an improved control circuit for a slit radiographic apparatus wherein the slit radiographic apparatus is comprised of an X-ray source adapted to scan a body to be radiographed through a slit diaphragm with a planar fan beam wherein the slit diaphragm is divided into a plurality of attenuating sections each associated with a controllable X-ray attenuating member and wherein there is provided a detection assembly for detecting the intensity of the radiation having passed through the body and divided into sections corresponding to attenuating sections of the slit diaphragm wherein each X-ray attenuating member is controlled by the control circuit including an energizing circuit and a comparison circuit for each attenuating member of the slit diaphragm whereby the comparison circuit compares an output signal of an associated section of the detection means to a reference signal to form an output coupled to the energizing circuit to control the attenuating member of a respective attenuating section of the slit diaphragm so that the difference between the output signal of the section of the detection assembly and the reference signal continously pursues zero value.
	abstract	Improvements in the fabrication of integrated circuits are driven by the decrease of the size of the features printed on the wafers. Current lithography techniques limits have been extended through the use of phase-shifting masks, off-axis illumination, and proximity effect correction. More recently, liquid immersion lithography has been proposed as a way to extend even further the limits of optical lithography. This invention described a methodology based on contact or proximity printing using a projection lens to define the image of the mask onto the wafer. As the imaging is performed in a solid material, larger refractive indices can be obtained and the resolution of the imaging system can be increased.
	claims	1. A method for oxidative erosion or for decontamination of a metallic surface, comprising:polarizing the metallic surface to be eroded or decontaminated by chemical etching using oxidative erosion; andplacing said polarized metallic surface in contact with an electrolytic solution containing an oxidant at a more anodic electric potential than a corrosion potential of said polarized metallic surface whereby electrolysis of the electrolytic solution is prevented, the electrolytic solution containing the oxidant including manganese VII,whereinsaid polarization is intermittent and is generated by at least one electric pulse;an anodic overpotential between an electric potential at which the metallic surface is polarized and the corrosion potential of the polarized metallic surface is between 0.005 and 0.800 V; andsaid method is conducted in the presence of ozone; andcurrent densities on the metallic surface to be eroded or decontaminated lie between 0.5 and 5.0 A·m−2. 2. The method according to claim 1, wherein the anodic overpotential between the electric potential at which the metallic surface is polarized and the corrosion potential of said surface is between 0.010 and 0.500 V. 3. The method according to claim 2, wherein the anodic overpotential between the electric potential at which the metallic surface is polarized and the corrosion potential of said surface is between 0.020 and 0.200 V. 4. The method according to claim 3, wherein the anodic overpotential between the electric potential at which the metallic surface is polarized and the corrosion potential of said surface is between 0.050 and 0.100 V. 5. The method according to claim 1, wherein a duration of each said at least one electric pulse is between about 1 sec and about 1 h. 6. The method according to claim 5, wherein the duration of each said at least one electric pulse is between about 10 sec and about 45 min. 7. The method according to claim 6, wherein the duration of each said at least one electric pulse is between about 1 min and about 30 min. 8. The method according to claim 7, wherein the duration of each said at least one electric pulse is between about 100 sec and about 1000 sec. 9. The method according to claim 1,wherein said at least one electric pulse comprises multiple electric pulses, andwherein a frequency of said multiple electric pulses ranges from 250 h−1 to 0.05 h−1. 10. The method according to claim 9, wherein said frequency of said multiple electric pulses ranges from 100 h−1 to 0.1 h−1. 11. The method according to claim 10, wherein said frequency of said multiple electric pulses ranges from 50 h−1 to 0.5 h−1. 12. The method according to claim 1, wherein said electrolytic solution contains nitric acid. 13. The method according to claim 1, wherein the manganese is initially added to said solution in the form of manganese II, manganese IV, manganese VII, or a mixture thereof. 14. The method according to claim 13, wherein the manganese is initially added to said solution at a concentration of less than 500 mg/L. 15. The method according to claim 14, wherein the manganese is initially added to said solution at a concentration of between 10 and 400 mg/L. 16. The method according to claim 15, wherein the manganese is initially added to said solution at a concentration of between 20 and 200 mg/L. 17. The method according to claim 16, wherein the manganese is initially added to said solution at a concentration of between 50 and 100 mg/L. 18. The method according to claim 1, further comprising:stabilizing manganese VII to manganese II. 19. The method according to claim 18, wherein said stabilizing comprises adding oxygenated water (H2O2) to said solution containing manganese VII. 20. The method for oxidative erosion or for decontamination of a metallic surface according to claim 1, wherein said method further comprises subjecting the metallic surface to be eroded or decontaminated to at least one non-corrosive rinsing, prior to said polarizing.
052664945	claims	1. A method of characterizing contaminated soil and determining an effective treatment approach for removing contaminants from said soil, said method comprising the steps of: (a) obtaining a representative contaminated soil sample from a site containing said contaminated soil; (b) (i) identifying particle size ranges from said contaminated soil by passing at least a portion of said representative contaminated soil sample through a series of particle size classifying means and (ii) analyzing each size range of particles for contaminants of interest in order to correlate the levels of said contaminants of interest with said size ranges to determine which of said size ranges has a greater proportion of said contaminants than other particle size ranges; (c) utilizing the information with respect to particle size ranges and contaminants from step (b)(ii), identifying a first extractant to be used to remove said contaminants of interest from said contaminated soil by passing at least a portion of said representative contaminated soil sample containing size ranges determined to have a greater proportion of said contaminants through a first bench scale soil washing process adapted to substantially correspond to a full-scale soil washing process for said contaminated soil, said first bench scale soil washing process adapted for removing said contaminants from contaminated soils having particle size ranges corresponding to said ranges determined in step (b)(ii), said first bench scale soil washing process including washing a known quantity of said representative soil sample with said first extractant, analyzing said washed sample for the quantity of said contaminants of interest remaining on said washed sample and (d) repeating step (c) with a second bench scale soil washing process on at least one additional representative contaminated soil sample obtained from said site with a second extractant, comparing the quantity of said contaminants remaining on said washed soil following said second bench scale soil washing process with the quantity of said contaminants remaining on said washed soil following said first bench scale soil washing process in order to determine which of said first or second bench scale soil washing processes favors extraction of said contaminant from said contaminated soil; and (e) identifying a suitable leachate treatment process for treating contaminants removed in step (c) from said representative contaminated soil sample. (a) drying any solids collected on said screens; (b) weighing said dried solids; (c) filtering all said rinse water to recover said fines; (d) drying said fines; and (e) analyzing said dried solids and fines for contaminants of interest. (a) selecting a first particle classifying means capable of isolating a particle size from said representative contaminated soil sample corresponding to a size identified in step (b); (b) placing said particle classifying means in communication with a receiver means; (c) placing a known quantity of said representative contaminated soil sample on said first particle classifying means and spraying said sample with a known quantity of extractant solution; (d) rinsing said sample with a known amount of rinse solution comprising water or extractant solution while collecting a mixture of rinse solution and fines in said receiver means and collecting a first product sample in said first particle classifying means; (e) drying and weighing said collected first product sample; (f) analyzing said dried first product sample for a contaminant of interest; (g) processing said mixture of rinse solution and fines to extract at least one contaminant of interest from said fines; (h) passing said fines through a second particle classifying means while rinsing said fines with a known quantity of rinse solution and collecting a second mixture comprising fines and rinse solution, said second particle classifying means classifying a smaller size of fines than said first particle classifying means; (i) collecting a second product sample in said second particle classifying means and drying and weighing said second product sample; (j) analyzing said dried second product sample for a contaminant of interest. 2. The method of claim 1 wherein said contaminants in said contaminated soil comprise heavy metals, radioactive compounds, organics, or a combination thereof. 3. The method of claim 1 wherein said particle size classifying means comprise a set of screens including a first screen and at least one additional screen, arranged in order of decreasing screen size relative to the direction of passing said representative contaminated soil sample through said screens. 4. The method of claim 3 wherein said screens comprise a range of sizes from No. 10 to No. 325. 5. The method of claim 3 wherein a known quantity of said representative contaminated soil sample is placed on said first screen and said first screen is rinsed with water to wash fines through said screen until no further fines removal is observed, and each successive said screen is rinsed with additional water until no further fines removal is observed, and step (b) further comprises the steps of: 6. The method of claim 1, wherein said bench scale soil washing process of step (c) comprises the steps of: 7. The process of claim 1 wherein the treatment process of step (d) is performed on contaminant-bearing media selected from the group wash liquor, leachate, fines and mixtures thereof. 8. The process of claim 1 wherein the treatment process of step (d) is selected from the group pH adjustment, precipitation, flocculation, filtration, settling, absorption and combinations thereof. 9. The process of claim 6 wherein said first and second particle classifying means comprise screens for sieving said representative contaminated soil sample.
	summary
	summary
055457967	claims	1. An article of manufacture, consisting essentially of: (1) a waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, and (2) concrete binder forming a matrix for the waste material, to provide the article, where the article is itself useful to isolate additional material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, where the article is made of different sized particles to provide high interior void volume filling, and the waste material used to make the article is fixed in the concrete matrix so that leaching of the waste material used to make the article is controlled. (1) a waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, and (2) concrete binder forming a matrix for the waste material, to provide the article, where the article is a transportable container the waste is in small discrete form, and the container is made of different sized particles to provide high interior void volume filling. (A) providing a contaminated material selected from at least one of: (B) mixing thoroughly: (C) forming the composition into a unitary, solid containment system which contains contaminated material, and binder acting as a matrix for the contaminated material, where the system contains different sized particles to provide high interior void volume filling; and (D) placing the containment system in direct or indirect contact with, radioactive, hazardous, or mixed waste. (A) providing quantities of radioactive material selected from the group consisting of radioactive metal, radioactive concrete, radioactive sand, radioactive gravel, radioactive plastic, radioactive liquid, and mixtures thereof, (B) processing the radioactive material without dilution with any more than about 15 weight % of non-radioactive material, to provide at least one of: (C) mixing (D) forming the composition into a unitary solid article, where the article contains difference sized particles to provide high interior void volume filling. (A) providing radioactive material selected from the group consisting of radioactive metal, radioactive concrete, radioactive sand, radioactive gravel, radioactive plastic, and mixtures thereof; and then (B) processing the radioactive material without dilution with any more than about 15 weight % of non-radioactive material, to provide at least one of (C) mixing (D) forming the composition into a containment system, where the containment system contains different sized particles to provide high interior void volume; and (E) curing the system into a unitary solid mass. (1) a waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, and (2) concrete binder forming a matrix for the waste material, where from 2 parts to 570 parts of waste material is used per 100 parts of concrete matrix, the waste material is fixed in the concrete matrix so that leaching of the waste material is controlled, the article is made of different sized particles to provide an article with high interior void volume filling, the article has a density over about 90% of theoretical density, and low permeability to water, and the waste material is in nonagglomerate form and uniformly and homogeneously dispersed, where the article is itself useful for a variety of functions in an area of waste environment. (1) a waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, and (2) concrete binder forming a matrix for the waste material, to provide the article, where the article itself is useful to isolate additional material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, and any radioactive waste used in the concrete matrix of the article itself can have cobalt-60 equivalents over 130 Bq/g, and is not diluted with any more than 15 weight % of nonradioactive material. (1) a waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, and (2) concrete binder forming a matrix for the waste material, where from 2 parts to 570 parts of waste material is used per 100 parts of concrete matrix, where the waste material is fixed in the concrete matrix so that leaching of the waste material is controlled, the article contains different sized particles to provide high interior void volume filling, and low permeability to water, the waste material is in non-agglomerate form and uniformly and homogeneously dispersed, and the article is itself useful for a variety of functions in an area of waste environment. 2. An article of manufacture, consisting essentially of: 3. A containment system comprising a concrete structure, the structure containing, as an actual part of its walls, radioactive metal, in the form of discrete fibers constituting from 2 weight % to 55 weight % of the structure, where the structure contains different sized particles to provide high interior void volume filling, and where the concrete structure itself is useful to isolate waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof. 4. The containment system of claim 3, where the structure is in the form of a transportable container, where a plastic sheet material covers at least one of the inside of the container and the outside of the container, where said plastic sheet material is closely attached to the container, the fibers have lengths from 0.5 cm to about 20 cm, and the container is placed in direct or indirect contact with contaminated material. 5. An article of manufacture, comprising a structure containing a series of different sized particles to provide high interior void volume filling, where at least one fine particulate selected from the group consisting of silica fume and flyash particles and mixtures thereof is close packed between larger particulate comprising cement, such that the structure has a high density, and where the structure also contains additives distributed therethrough, selected from the group consisting of uniformly dispersed bars, fibers, generally spherical particles and amorphous particles, and mixtures thereof, where at least one of the additives is radioactive waste or hazardous waste. 6. The article of claim 5, where the structure is a containment system, where the larger particulate may also include filler, and aggregate particles and mixtures thereof, the bars have lengths from about 25 cm to about 150 cm and diameters of from about 0.10 cm to about 3 cm, the fibers have lengths from about 0.5 cm to about 20 cm and a length:width aspect ratio of between 200:1 and 20:1, the generally spherical particles have diameters from about 0.001 mm to about 30 mm, and the amorphous particles have a thickness of from 0.01 mm to 30 mm. 7. The article of claim 5, where the additive is radioactive metal fibers. 8. The article of claim 5, where the additive is generally spherical radioactive particles. 9. The article of claim 5, in container form, where the additives are selected from at least one of metal bars, radioactive metal fibers, and generally spherical radioactive concrete particles, and where a closely attached plastic sheet material covers at least one of the inside of the container or the outside of the container. 10. A method of mixing radioactive or hazardous waste into a binder matrix to form a containment system comprising the steps of: 11. The method of claim 10, where the binder material is a concrete mixture of cement, sand, aggregate and water, the contaminated material is selected from at least one of radioactive bars, radioactive fibers, generally spherical radioactive particles, amorphous radioactive particles and stabilized radioactive liquid, the cured containment system has a density over about 90% of theoretical density, and low permeability to water, where 2 parts to 570 parts of contaminated material can be used per 100 parts of binder material and any radioactive material used in the binder matrix itself is in non-agglomerate form and uniformly and homogeneously dispersed, and can have cobalt-60 equivalents over 130 Bq/g, and where the contaminated material is fixed in the binder matrix so that leaching of the contaminated material is controlled. 12. A method of making an article utilizing radioactive, hazardous, or mixed waste as a component of the structure, comprising the steps: 13. The method of claim 12, where about 1 to about 25 parts by weight of hazardous waste material selected from at least one of toxic chemicals, plastics, and soil is mixed in step (C) and the binder is a concrete mixture. 14. The method of claim 12, where no hazardous waste material is added in step (C). 15. The method of claim 12, where the system is cured in step (D) into a containment system which contains at least radioactive material, and uncontaminated binder acting as a matrix for the radioactive material. 16. A method of making a containment system utilizing radioactive waste as a component of the system, comprising the steps: 17. The method of claim 16, where the radioactive material is radioactive metal which is processed into fibers in step (D) the composition is formed into a container, and as a last step the system is cured into a unitary solid mass. 18. The method of claim 16, where the radioactive material is radioactive concrete which is processed into generally spherical particles and in step (D) the composition is formed into a container. 19. The method of claim 16, where the concrete mixture has a consistency of from 3 to 7 cm slump and the plasticizer added increases the consistency to from 12 to 17 cm slump. 20. The method of claim 16, where the cured containment system has a density over about 90% of theoretical density, and low permeability to water, where 2 parts to 570 parts of processed radioactive material can be used per 100 parts of cement material and any radioactive material used is in non-agglomerate form and uniformly and homogeneously dispersed, and can have cobalt-60 equivalents over 130 Bq/g, and where the radioactive material is fixed in the binder matrix so that leaching of the radioactive material is controlled. 21. A container made by the method of claim 16. 22. A container comprising concrete and from 2 weight % to 55 weight % contaminated metal fibers having lengths from about 0.5 cm to about 20 cm and a length:width aspect ratio of between 200:1 and 20:1, where the container contains different sized particles to provide high interior void volume filling and a density over about 90% of theoretical density. 23. The container of claim 22, where a closely attached plastic sheet material covers at least one of the inside of the container and the outside of the container. 24. The container of claim 22, being transported with waste contained therein to a storage location and stored. 25. The container of claim 22 where the metal fibers constitute from about 2 weight % to about 30 weight % of the container. 26. An article of manufacture, consisting essentially of: 27. An article of manufacture, consisting essentially of: 28. An article of manufacture, consisting essentially of: 29. The containment system of claim 3, where the structure has a density over about 90% of theoretical density and low permeability to water, the metal is in non-agglomerate form and uniformly and homogeneously dispersed, and is fixed in the concrete matrix so that leaching of the radioactive metal is controlled, and any radioactive metal used in the structure walls can have cobalt-60 equivalents over 130 Bq/g, and is not diluted with any more than 15 weight % of non-radioactive material. 30. The article of claim 5, where the structure itself is useful to isolate waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, where any radioactive waste additive used in the structure itself can have cobalt-60 equivalents over 130 Bq/g, and is not diluted with any more than 15 weight % of non-radioactive material, and where the structure has a density over about 90% of theoretical density, and low permeability to water, and where any radioactive or hazardous waste is in non-agglomerate form and uniformly and homogeneously dispersed, and is fixed in the structure so that leaching of the radioactive or hazardous waste is controlled. 31. The container of claim 22, where the container itself is useful to isolate waste material selected from the group consisting of radioactive waste, hazardous waste, and mixtures thereof, where the contaminated metal fibers are in non-agglomerate form and uniformly and homogeneously dispersed, where any contaminated metal fibers used in the container itself can have cobalt-60 equivalents over 130 Bq/g, and are not diluted with any more than 15 weight % of non-radioactive material, and where the container has low permeability to water. 32. The method of claim 12, where the cured article has a density over about 90% of theoretical density, and low permeability to water, and any radioactive material used is in non-agglomerate form and uniformly and homogeneously dispersed and can have cobalt-60 equivalents over 130 Bq/g, and where the radioactive and hazardous waste material is fixed in the binder so that leaching of such material is controlled.
	abstract	Disclosed herein are systems and methods for a modular reconfigurable shielding system for one or more storage containers in temporary or long term storage. The system comprises shield panels which may be used to shield external faces of containers in a storage configuration to reduce the overall amount of shielding required in a storage facility. Reducing the amount of shielding reduces the storage footprint of each container thus increasing storage capacity and efficiency of the storage facility. The modularity of the shield panels allows storage containers to be easily added and removed from the storage configuration. Additionally, modular shielding allows the amount and type of shielding to be easily reconfigured for differing requirements and storage contents.
	abstract	The assembly apparatus comprises a frame for supporting the straps of the first set of straps in mutually parallel positions, clamping and engagement for clamping on each of the straps of a second set of straps in succession and for engaging them with the straps of the first set in position in the support frame, and at least one comb that is movable between a disengaged position and a position in which it engages each of the straps of the first set.
045132046	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In FIGS. 1-7, a holder, container or housing containing or carrying a radioactive source 2 is illustrated. The source 2 comprises a radioactive isotope which is used in a diagnostic motion correction scheme. Thus, the housing may be termed a dual isotope motion correction centroid point source holder. In the present case, the source 2 is commercially available for nuclear medicine purposes. In particular, it is a gamma radiation source available from The Radiochemical Centre, Amersham, England. A 2 mm diameter spherical point source 4 is located at one end of a closed-end cylindrical stainless steel housing 6 of 3 mm diameter and 10 mm length. The other end of the source housing 6 is plugged with an approx. 8 mm long stainless steel plug. The point source 4 is a pellet or bead containing the radioactive isotope. For motion correction applications, preferably an americium source is used. The radioactive material americium 241 has a comparatively low energy with regard to the emitted gamma radiation. This energy differs markedly from the energy of gamma radiation emitted by a tracer which is conventionally administered to a patient in nuclear medicine. The half-life of americium is several hundred years. This has the advantage that during patient examination and during the lifetime of a scintillation camera, a replacement of the radiation source 2 is not necessary. An americium point source can also be used as a marker in nuclear imaging by means of a scintillation camera such as the Anger camera. The housing contains a lower or bottom part 8 and an upper or top part 10. Both parts or sections 8 and 10 are made of a shielding material, preferably of a metal. A compound containing tungsten is preferred. A tungsten alloy known as "Mallory 1000" has been found especially useful. The holder sections 8 and 10 are of a material containing tungsten rather than lead because of its durability and its lighter weight. However, also other radiation attenuating materials may be used. The bottom part 8 represents a first radiation shielding body. As can be seen in FIG. 1, it contains at its upper end a protrusion 12. The protrusion 12 and the base portion 14 of the bottom part 8 are both of cylindrical or disk shape. The protrusion 12 contains a first recess or chamber 16 for holding the cylindrical source housing 6 containing the radioactive americium source 4 therein. In particular, the first recess 16 is a channel extending from the periphery of the cylindrical protrusion 12 to its middle section. Here it merges into another channel or chamber 18 which is arranged perpendicularly thereto. The channels 16 and 18 are shaped so that the point source 4 becomes positioned in the center of the disk structure 8. Both channels 16, 18 extend from the upper face end of the protrusion 12 to the upper end of the base portion 14. In operation, the source 2 is kept in the channel 16 by means of a cement. The top part 10 represents a second radiation shielding body. On its lower end it contains a second recess 20 which is illustrated in FIG. 1 in broken lines. The recess 20 as well as the top part 10 have a cylindrical or disk shape. The dimension of the second recess 20 is such that the protrusion 12 of the bottom part 8 fits tightly therein. In the upper surface of the upper part 10 is provided a conical recess 22. The conical angle of this recess 22 is designated by .alpha.. The recess 22 merges into a central aperture 23 connecting the upper recess 22 with the lower recess 20. The diameter d of the central aperture 23 is preferably d=2 mm, see FIGS. 4 and 5. The aperture 23 and the recess 22 are provided for permitting the passage of radiation quanta from the point source 4 at angles between 0.degree. and 60.degree. towards the upper surface. To the lower end of the lower part 8, there may be attached a label 25, see FIG. 3. This label 25 contains information such as radiation identification and a caution note. There are also provided means for connecting the top part 10 to the bottom part 8. These means comprise two openings 24 extending axially through the bottom part 8 and two openings 26 extending axially through the top part 10. These openings 24, 26 are provided at the rim portions of the parts 8 and 10, respectively, and they are aligned with respect to each other. The connecting means may further comprise rivets 28, extending through the openings 24 and 26 as illustrated in FIG. 3. Thus, in the preferred embodiment the parts 8 and 10 are kept together by the rivets 28 running through boxes at diagonally opposite sides of the parts 8 and 10. During assembly of the structure, the radiation source 2 is inserted and cemented into the first recess 16 and subsequently the protrusion 12 is plugged into the second recess 20. Now the lower end face of the upper part 10 joins the upper end face of the bottom portion 14. Then the top part 10 is connected to the bottom part 8 by means of the aforementioned rivets 28. In the closed status of the parts 8 and 10, the radioactive source 2 will emit gamma radiation basically through the conical aperture 22. The centroid source holder may now be attached to the patient. The openings 22, 23 in the center of the upper part 10 serve as a port to pass the emitted radiation from the point source 4 towards the scintillation camera. After assembly, the upper and lower parts 8, 10 both enclose the source 2 in a tight manner. The point source 4 is completely encapsulated with the exception of the area where the aperture 23 and the conical recess 22 are provided. Due to this feature, an optimum shielding of the patient and of the technician or physician performing the examination is achieved. Examinations have proven that a wide range of conical angles .alpha. can be used. However, an angle of approximately .alpha.=120.degree. has proven to be best for certain applications. This angle .alpha. is especially useful if a movable scintillation camera is used. It ensures that the scintillation camera exerts a constant count rate at various positions of the camera with respect to the housing 8, 10. In order that the angle .alpha. can be used optimally, the height b of the protrusion 12 should not markedly exceed the diameter of the source 2. Thus, the 120.degree. angle is empirically chosen. It allows for angular placement of the camera head relative to the source. Yet it is not too wide an angle to prevent shielding of the radiation from the patient and surrounding environment (nurses, etc.). A vertical walled opening would not permit many counts to reach the camera head if the head is placed at an angle because the parallel collimator would not accept rays at angles to the axes of the collimator paths. The size of the container is small so that it can be easily taped, e.g. to the chest of a patient by a surgical tape. It will move with the movement of the patient's chest. In a preferred embodiment, the container dimensions are A=2.5 cm; B=1.9 cm; a=7 mm; b=3 mm; c=4 mm. While the forms of the housing for a radioactive source herein described constitute preferred embodiments of the invention, it is to be understood that the invention is not limited to these precise forms of assembly, and that a variety of changes may be made therein without departing from the scope of the invention.
061730262	summary	BACKGROUND OF THE INVENTION This invention relates generally to nuclear reactors and more particularly, to a system for monitoring reactor instability and controlling reactor suppression. In known types of nuclear reactors, such as boiling water reactors (BWR), the reactor core includes a plurality of fuel bundles arranged in an array capable of self-sustained nuclear fission reaction. The core is contained in a reactor pressure vessel (RPV). The typical core is submerged in a liquid such as water, which serves as both a core coolant and a neutron moderator. Each fuel assembly includes a flow channel through which water is pumped upwardly from a lower plenum to an upper plenum. A plurality of control rods containing neutron absorbing material are insertable between the fuel bundles to control the reactivity of the core. To monitor core conditions, it is common practice to distribute neutron detectors both radially and axially throughout the core. The signals from these neutron detectors are utilized to monitor the power density of the core and to initiate corrective actions, including reactor suppression, in the event of detected instability. A nuclear reactor operates under three distinctly different stability regimes. These regimes are a stable reactor state, a reactor instability threshold state and an unstable reactor state. Reactor instability occurs when fuel cladding heat flux and channel coolant flow rates deviate from steady state conditions during power oscillations significantly above the normal neutron noise level. Reactor instability must be monitored to prevent damage to the core and to within fuel safety limits and can be accomplished by either detecting and suppressing instability induced power oscillations, or preventing them altogether. Known "detect and suppress" systems are based on a common approach. Generally, neutron detectors, for example, local power range monitors (LPRMs), are placed within the core of the reactor. The neutron detectors generate signals indicative of reactor thermal-hydraulic oscillation frequencies. These signals are characteristic of power oscillations of the reactor. Some known "detect and suppress" systems monitor successive oscillation periods and provide a final oscillation amplitude trigger to generate a reactor trip signal. Another known system monitors of an oscillation growth rate limit and if the limit is exceeded, the system generates a reactor trip. Yet another known system compares the neutron detector signal to an amplitude trip setpoint and if the setpoint is exceeded, a reactor trip signal is generated. Although the known systems generally provide satisfactory results, such systems permit the development of significant power oscillations prior to actuation of the suppression function. As a result, rigorous analysis of minimum critical power ratio (MCPR) performance in the presence of core power and flow oscillations is necessary to prevent exceeding the fuel safety limits or damaging the core. In particular, the transient thermal-hydraulic behavior within the limiting fuel bundle must be related to neutron flux oscillations observed by various neutron detectors. Another shortcoming of known systems is the potential spatial effects related to the oscillation mode and its impact on the magnitude of the oscillation as observed by any given neutron detector. Combinations of neutron detector signals must be related to a reactor trip setpoint for each of the detection methods. As a result, the known systems are time consuming to determine the appropriate trip levels and allow significant power oscillations to occur prior to suppression. Additionally, current "detect and suppress" systems are still evolving since the quantification of MCPR performance as a function of power oscillation scenarios for the full spectrum of core designs and operating conditions is extremely challenging. It would be desirable to provide a system that, prior to the development of significant power oscillations, facilitates suppression of the nuclear reactor upon reaching the threshold of reactor instability. BRIEF SUMMARY OF THE INVENTION These and other objects are attained by a power oscillation monitoring system which, in one embodiment, facilitates monitoring a nuclear reactor for instability and generates a thermal-hydraulic instability signal facilitating reaction suppression. The present system discriminates instability threshold conditions from stable reactor operation. Particularly, at the threshold of reactor instability, a qualitative change in the reactor core neutronic/thermal-hydraulic response occurs, and the reactor realigns into a strongly coupled configuration, resulting in an increasingly coherent response. This change affects increasingly larger regions of the core and a number density of neutron detectors located in the core reach or exceed a target successive oscillation period confirmation count. Upon the number density exceeding a defined level, an instability signal is generated by the system. By monitoring the neutron detector output signals, the system can reliably and efficiently detect the transition to a coherent core response, which is characteristic of the reactor at the threshold of instability. In one form, the monitoring system is implemented in a computer workstation having a memory and a processor. The workstation is electrically coupled to neutron flux detectors, and a Period Based Algorithm (PBA) and a Confirmation Density Algorithm (CDA) are stored in the workstation memory. Under the control of the PBA algorithm, the system evaluates the periodicity of the neutron detector output signals to identify the presence of a density wave in the reactor. During stable reactor operation, the system occasionally generates a successive power oscillation period confirmations as a result of global perturbations to the steady state neutron flux. The CDA utilizes the PBA output signals to identify a confirmation density, which is the fraction of active neutron detectors in the core that reach a target successive oscillation period confirmation count. When the confirmation density exceeds an instability threshold setpoint, the instability signal generated by the CDA changes from a first state to a second state. Following recognition of this condition, the system provides protection of a fuel thermal safety limit by generating the instability signal prior to any growth in power oscillation amplitude. The instability signal facilitates use of the control rods to maintain the stability of the reactor. By providing suppression at these conditions, the development of significant power oscillations is avoided. The above described system facilitates maintaining a nuclear reactor in the stable state. The system avoids the characterization of MCPR performance as a function of growing power oscillations based on local neutron noise characteristics sensed by a few of the neutron detectors. In addition, the system eliminates the generation of spurious instability signals based on the thermal-hydraulic behavior of the reactor. The above-described system monitors the nuclear reactor for instability and facilitates suppression of the reactor prior to development of significant power oscillations. In addition, the system does not require rigorous analysis of MCPR performance.
	description	The present invention relates to a scintillator panel and a radiation detector. Patent Literature 1 to Patent Literature 3 are known as technologies in this field. Patent Literature 1 discloses a scintillator panel. The scintillator panel has a metal film provided between a resin substrate and a fluorescent body layer. Patent Literature 2 discloses a radiation detection apparatus including a scintillator panel. The scintillator panel has a scintillator layer having cesium iodide as a main component. Thallium is doped into the scintillator layer. The thallium is highly concentrated near an interface of the scintillator layer with respect to a substrate. According to a concentration distribution of the thallium, an optical output is improved. Patent Literature 3 discloses a radiation detector including a fluorescent body layer. The radiation detector has a scintillator layer having cesium iodide as a main component. Thallium is doped into the scintillator layer. The thallium is highly concentrated on a substrate side in the scintillator layer. According to a concentration distribution of the thallium, adhesion between a sensor substrate and the fluorescent body layer is improved. Patent Literature 1: PCT International Publication No. WO2011/065302 Patent Literature 2: Japanese Unexamined Patent Publication No. 2008-51793 Patent Literature 3: Japanese Unexamined Patent Publication No. 2012-98110 Growth substrates for growing a scintillator layer sometimes have moisture permeability of allowing moisture to permeate thereinto. Moisture which has peitneated into a growth substrate arrives at a base portion of the scintillator layer. It is known that a scintillator layer formed of cesium iodide is deliquescent. Due to moisture supplied through the growth substrate, deliquescence occurs in the base portion of the scintillator layer. As a result, characteristics of a scintillator panel deteriorate. Accordingly, in this field, it is desired that the moisture resistance of a scintillator panel having a scintillator layer formed of cesium iodide be improved. For example, a scintillator panel of Patent Literature 1 has a metal film provided between a substrate and a fluorescent body layer. The metal film hinders movement of moisture from the resin substrate to the fluorescent body layer. An object of the present invention is to provide a scintillator panel and a radiation detector, in which the moisture resistance can be improved. According to an aspect of the present invention, there is provided a scintillator panel including a substrate, an intermediate layer formed on the substrate and made of an organic material, a barrier layer formed on the intermediate layer and including thallium iodide as a main component, and a scintillator layer formed on the barrier layer and constituted of a plurality of columnar crystals including cesium iodide with thallium added thereto as a main component. The scintillator layer of the scintillator panel is formed on the substrate with the intermediate layer and the barrier layer therebetween. The barrier layer includes thallium iodide as a main component. The barrier layer has properties of allowing scarcely any moisture to permeate thereinto. As a result, moisture which tends to move from the intermediate layer made of an organic material to the scintillator layer can be blocked by the barrier layer. That is, since deliquescence in a base portion of the scintillator layer is curbed, deterioration in characteristics of the scintillator panel can be curbed consequently. Accordingly, it is possible improve the moisture resistance of the scintillator panel. In the scintillator panel, the substrate may include any one of a metal material, a carbon material, a glass material, and a resin material as a main component. According to this constitution, it is possible to apply characteristics based on material characteristics to the substrate. In the scintillator panel, the organic material may include any one of a xylylene resin, an acrylic resin, a silicone resin, and a polyimide or polyester resin. According to this constitution, it is possible to apply characteristics based on the material characteristics to the intermediate layer. According to another aspect of the present invention, there is provided a radiation detector including a scintillator panel having a substrate, an intermediate layer formed on the substrate and made of an organic material, a barrier layer formed on the intermediate layer and including thallium iodide as a main component, and a scintillator layer formed on the barrier layer and constituted of a plurality of columnar crystals including cesium iodide with thallium added thereto as a main component; and a sensor substrate including a photo-detection surface provided with a photoelectric conversion element receiving light generated in the scintillator panel. The photo-detection surface of the sensor substrate faces the scintillator layer. In the radiation detector, light is generated due to radiation incident on the scintillator panel, and light is detected by the photoelectric conversion element provided on the photo-detection surface. The radiation detector has the intermediate layer made of an organic material and the barrier layer including thallium iodide as a main component between the substrate and the scintillator layer. According to the barrier layer, movement of moisture from the intermediate layer to the scintillator layer can be blocked. Accordingly, since deliquescence in the base portion of the scintillator layer is curbed, deterioration in characteristics of the scintillator panel can be curbed. As a result, in the radiation detector, deterioration in characteristics of detecting radiation is curbed. Accordingly, it is possible for the radiation detector to have improved moisture resistance. In the radiation detector, the substrate may include any one of a metal material, a carbon material, a glass material, and a resin material as a main component. According to this constitution, it is possible to apply characteristics based on the material characteristics to the substrate. In the radiation detector, the organic material may include any one of a xylylene resin, an acrylic resin, a silicone resin, polyimide, and a polyester resin. According to this constitution, it is possible to apply characteristics based on the material characteristics to the intermediate layer. According to still another aspect of the present invention, there is provided a radiation detector including a substrate, an intermediate layer formed on the substrate and made of an organic material, a barrier layer formed on the intermediate layer and including thallium iodide as a main component, and a scintillator layer formed on the barrier layer and constituted of a plurality of columnar crystals including cesium iodide with thallium added thereto as a main component. The substrate has a photo-detection surface provided with a photoelectric conversion element receiving light generated in the scintillator layer. In the radiation detector, light is generated due to radiation incident on the scintillator panel. Light is detected by the photoelectric conversion element provided on the photo-detection surface. The radiation detector has the intermediate layer made of an organic material and the barrier layer including thallium iodide as a main component between the substrate and the scintillator layer. According to the barrier layer, movement of moisture from the intermediate layer to the scintillator layer can be blocked. Accordingly, since deliquescence in the base portion of the scintillator layer is curbed, deterioration in characteristics of the scintillator panel can be curbed. As a result, in the radiation detector, deterioration in characteristics of detecting radiation is curbed. Accordingly, it is possible for the radiation detector to have improved moisture resistance. In the radiation detector, the substrate may include any one of a metal material, a carbon material, and a glass material as a main component. According to this constitution, it is possible to apply characteristics based on the material characteristics to the substrate. In the radiation detector, the organic material may include any one of a xylylene resin, an acrylic resin, a silicone resin, a polyimide or polyester resin, a siloxane resin, and an epoxy resin. According to this constitution, it is possible to apply characteristics based on the material characteristics the intermediate layer. According to the present invention, there are provided a scintillator panel and a radiation detector, in which the moisture resistance can be improved. Hereinafter, with reference to the accompanying drawings, embodiments of the present invention will be described in detail. In description of the drawings, the same reference signs will be applied to the same elements, and duplicate description will be omitted. As illustrated in FIG. 1, a scintillator panel 1 according to a first embodiment has a substrate 2, a resin protective layer 5 (intermediate layer), a scintillator layer 4, and a protective film 6. The scintillator panel 1 is combined with a photoelectric conversion element (not illustrated) and is used as a radiation image sensor. The substrate 2, the resin protective layer 5, a barrier layer 3, and the scintillator layer 4 are laminated in this order in a thickness direction thereof and constitute a laminated body 7. Specifically, the resin protective layer 5 is formed on the substrate 2. The barrier layer 3 is formed on the resin protective layer 5. The scintillator layer 4 is formed on the barrier layer 3. The resin protective layer 5 and the barrier layer 3 are present between the substrate 2 and the scintillator layer 4. The substrate 2 and the scintillator layer 4 do not directly come into contact with each other. The laminated body 7 is covered with the protective film 6. The substrate 2 constitutes a base body of the scintillator panel 1. The substrate 2 exhibits a rectangular shape, a polygonal shape, or a circular shape in a plan view. The thickness of the scintillator panel 1 is within a range of 10 micrometers to 5,000 micrometers. As an example, the thickness of the scintillator panel 1 is 100 micrometers. The substrate 2 has a substrate front surface 2a, a substrate rear surface 2b, and a substrate side surface 2c. The substrate 2 is made of a metal material, a carbon material, a ceramic material, or a resin material. Examples of a metal material include aluminum, stainless steel (SUS), and copper. Examples of a carbon material include amorphous carbon. Examples of a ceramic material include glass and alumina. Examples of a resin material include polyethylene terephthalate, polyethylene naphthalate, polyimide, and polyetheretherketone. When the substrate 2 is made of a metal material, the resin protective layer 5 hinders the scintillator layer 4 from coming into direct contact with the substrate 2. As a result, the substrate 2 does not come into direct contact with the scintillator layer 4, and therefore corrosion of the metal substrate 2 due to direct contact is prevented. Accordingly, the resin protective layer 5 has an area larger than at least the scintillator layer 4. Alternatively, when the substrate 2 is made of a carbon material, a ceramic material, or a resin material, film formation of a root part of the scintillator layer 4 constituted of a plurality of columnar crystals is favorably performed with respect to the substrate 2 by providing the resin protective layer 5 on the substrate 2. As a result, favorable columnar crystals can grow. The resin protective layer 5 has a protective layer front surface 5a, a protective layer rear surface 5b, and a protective layer side surface 5c. The protective layer front surface 5a faces the barrier layer 3. The protective layer rear surface 5b faces the substrate front surface 2a. The protective layer side surface 5c is flush with the substrate side surface 2c. The resin protective layer 5 is formed on the entire surface of the substrate front surface 2a. The resin protective layer 5 may cover the substrate rear surface 2b and the substrate side surface 2c from above, in addition to the substrate front surface 2a. The resin protective layer 5 may cover the entirety of the substrate 2. The resin protective layer 5 is made of a resin material. Examples of a resin material include a xylylene resin such as polyparaxylylene, an acrylic resin, a silicone resin, and a polyimide or polyester resin. The barrier layer 3 hinders movement of moisture from the resin protective layer 5 to the scintillator layer 4. The barrier layer 3 is formed on a region of a part on the protective layer front surface 5a. When viewed in the thickness direction, the barrier layer 3 is smaller than the resin protective layer 5 and the substrate 2. The thickness of the barrier layer 3 is within a range of 0.001 micrometers to 1.0 micrometer. As an example, the thickness of the barrier layer 3 is 0.06 micrometers (600 angstroms). The barrier layer 3 has a barrier layer front surface 3a, a barrier layer rear surface 3b, and a barrier layer side surface 3c. The barrier layer front surface 3a faces the scintillator layer 4. The barrier layer rear surface 3b faces the protective layer front surface 5a. The barrier layer 3 includes thallium iodide (TlI) as a main component. For example, the TEE content of the barrier layer 3 may be within a range of 90% to 100%. When the TlI content in the barrier layer 3 is 90% or more, it may be stated that the barrier layer 3 has TlI as a main component. For example, the barrier layer 3 may be formed by a two-source vapor deposition method. Specifically, a first vapor deposition source containing cesium iodide (CsI) and a second vapor deposition source containing thallium iodide (TlI) are utilized. The barrier layer 3 is formed by performing vapor deposition of TlI on a substrate prior to CsI. As an example, the thickness of the barrier layer 3 is approximately 600 angstroms. The thickness of the barrier layer 3 can be measured by causing a scintillator layer and a substrate to peel off using a strong adhesive tape or the like and analyzing a substrate interface using an X-ray fluorescence analysis (XRF) apparatus. Examples of X-ray fluorescence analysis apparatuses can include ZSX Primus of RIGAKU Corporation. The scintillator layer 4 receives radiation and generates light corresponding to the radiation. The thickness of the scintillator layer 4 is within a range of 10 micrometers to 3,000 micrometers. As an example, the thickness of the scintillator layer 4 is 600 micrometers. The scintillator layer 4 is a fluorescent body material and including cesium iodide with thallium added thereto as a main component. The cesium iodide includes thallium as a dopant (CsI:Tl). For example, the CsI content of the scintillator layer 4 may be within a range of 90% to 100%. When the CsI content of the scintillator layer 4 is 90% or more, it may be stated that the scintillator layer 4 has CsI as a main component. The scintillator layer 4 is constituted of a plurality of columnar crystals. Each of the columnar crystals exhibits a light guiding effect. Accordingly, the scintillator layer 4 is suitable for high-resolution imaging. The scintillator layer 4 may be formed by a vapor deposition method, for example. The scintillator layer 4 has a scintillator layer front surface 4a, a scintillator layer rear surface 4b, and a scintillator layer side surface 4c. The scintillator layer 4 is formed on the barrier layer 3 such that the scintillator layer rear surface 4b faces the barrier layer front surface 3a. The barrier layer 3 is present between the scintillator layer 4 and the resin protective layer 5. The scintillator layer 4 does not directly come into contact with the resin protective layer 5. When viewed in the thickness direction, the barrier layer 3 is smaller than the substrate 2 and the resin protective layer 5. Similarly, in the scintillator layer 4 as well, when viewed in the thickness direction, the scintillator layer 4 is smaller than the substrate 2 and the resin protective layer 5. The scintillator layer 4 includes a plurality of columnar crystals extending in the thickness direction of the scintillator layer 4. Base portions of the plurality of columnar crystals constitute the scintillator layer rear surface 4b. The base portions come into contact with the barrier layer front surface 3a of the barrier layer 3. Tip portions of the plurality of columnar crystals constitute the scintillator layer front surface 4a. The scintillator layer 4 exhibits a truncated pyramid shape. The scintillator layer side surface 4c is tilted with respect to the thickness direction of the scintillator layer side surface 4c. The scintillator layer side surface 4c is a slope (inclination). Specifically, when the scintillator layer 4 is viewed in a cross-sectional view in a direction orthogonal to the thickness direction, a cross section exhibits a trapezoidal shape. That is, one side on the scintillator layer front surface 4a side is shorter than one side on the scintillator layer rear surface 4b side. The protective film 6 covers the laminated body 7. As a result, the protective film 6 protects the laminated body 7 from moisture. The protective film 6 covers the substrate rear surface 2b, the substrate side surface 2c, the protective layer side surface 5c, the barrier layer side surface 3c, the scintillator layer side surface 4c, and the scintillator layer front surface 4a. The thickness of the protective film 6 may be substantially the same at all places where it is formed. In addition, the thickness of the protective film 6 may vary at every place. In the protective film 6, for example, a film portion formed on the scintillator layer front surface 4a is thicker than film portions formed on the substrate rear surface 2b, the substrate side surface 2c, the barrier layer side surface 3c, and the scintillator layer side surface 4c. The protective film 6 may include polyparaxylylene as a main component. The protective film 6 may be formed by a chemical vapor deposition (CVD) method, for example. In the scintillator panel 1, the barrier layer 3 is provided between the resin protective layer 5 and the scintillator layer 4. The barrier layer 3 includes thallium iodide as a main component. The barrier layer 3 has properties of allowing scarcely any moisture to permeate thereinto. Accordingly, moisture which tends to move from the resin protective layer 5 to the scintillator layer 4 can be blocked by the barrier layer 3. As a result, deliquescence in the base portion of the scintillator layer 4 is curbed. Therefore, deterioration in characteristics of the scintillator panel 1 can be curbed. Accordingly, the moisture resistance of the scintillator panel 1 can be improved. A radiation detector according to a second embodiment will be described. Actually, a region (side) for achieving electrical conduction is provided on a sensor panel 11. However, it is not illustrated in each of the drawings for the sake of convenience. As illustrated in FIG. 2, a radiation detector 10 has the sensor panel 11 (sensor substrate), the resin protective layer 5, the barrier layer 3, the scintillator layer 4, and a sealing portion 12. Radiation received from a sealing plate 14 is incident on the scintillator layer 4. The scintillator layer 4 generates light corresponding to the radiation. The light passes through the resin protective layer 5 and the barrier layer 3 and is incident on the sensor panel 11. The sensor panel 11 generates an electrical signal in response to the incident light. The electrical signal is output through a predetermined electric circuit. According to the electrical signal, a radiation image is obtained. The sensor panel 11 has a panel front surface 11a, a panel rear surface 11b, and a panel side surface 11c. The sensor panel 11 is a CCD sensor, a CMOS sensor, or a TFT panel having a photoelectric conversion element 16. The sensor panel 11 includes a semiconductor such as silicon, or glass as a main component. The sensor panel 11 may include an organic material as a main component. Examples of an organic material include polyethylene terephthalate (PET), polyethylene naphthalate (PEN), and polyimide (PI). A plurality of photoelectric conversion elements 16 are disposed on the panel front surface 11a in a two-dimensional manner. A region on the panel front surface 11a on which a plurality of photoelectric conversion elements 16 are disposed is a photo-detection region S1 (photo-detection surface). In addition to the photo-detection region S1, the panel front surface 11a includes a surrounding region S2 surrounding the photo-detection region S1. In order to protect the photoelectric conversion elements 16, the resin protective layer 5 is provided. The resin protective layer 5 is a polyimide or siloxane resin or an epoxy resin, for example. In order to enhance crystallinity of the scintillator layer made of a plurality of columnar crystals, the resin protective layer 5 similar to that in the first embodiment may be provided. The sealing portion 12 covers a portion of the protective layer rear surface 5b of the sensor panel 11, the barrier layer 3, and the scintillator layer 4. The sealing portion 12 is fixed to the surrounding region S2 on the protective layer rear surface 5b. The sealing portion 12 air-tightly maintains an internal space formed by the sealing portion 12 and the resin protective layer 5. Due to this constitution, the scintillator layer 4 is protected from moisture. The sealing portion 12 has a sealing frame 13 and the sealing plate 14. The sealing frame 13 has a frame front surface 13a, a frame rear surface 13b, and a frame wall portion 13c. The frame wall portion 13c joins the frame front surface 13a and the frame rear surface 13b to each other. The height of the frame wall portion 13c (that is, the length from the frame front surface 13a to the frame rear surface 13b) is higher than the height from the protective layer rear surface 5b to the scintillator layer rear surface 4b. A gap is formed between the scintillator layer rear surface 4b and the sealing plate 14. The sealing frame 13 may be constituted of a resin material, a metal material, or a ceramic material, for example. The sealing frame 13 may be solid or hollow. The frame front surface 13a and a plate rear surface 14b, and the frame rear surface 13b and the resin protective layer 5 may be joined to each other using an adhesive. The sealing plate 14 is a plate material having a rectangular shape in a plan view. The sealing plate 14 has a plate front surface 14a, the plate rear surface 14b, and a plate side surface 14c. The plate rear surface 14b is fixed to the frame front surface 13a. The plate side surface 14c may be flush with an outer surface of the frame wall portion 13c. The sealing plate 14 may be constituted of a glass material, a metal material, a carbon material, or a barrier film, for example. Examples of a metal material include aluminum. Examples of a carbon material include CFRP. Examples of a barrier film include a laminated body of an organic material layer (PET or PEN) and an inorganic material layer (SiN). In the radiation detector 10, light is generated due to radiation incident on the scintillator layer 4. Light is detected by the photoelectric conversion elements 16 provided in the photo-detection region S1. The radiation detector 10 has the barrier layer 3 including thallium iodide as a main component between the resin protective layer 5 and the scintillator layer 4. The barrier layer 3 blocks movement of moisture from the resin protective layer 5 to the scintillator layer 4. Accordingly, deliquescence in the base portion of the scintillator layer 4 is curbed. As a result, in the radiation detector 10, it is possible to curb deterioration in detection characteristics. Hereinabove, embodiments of the present invention have been described. However, the present invention is not limited to the foregoing embodiments and can be performed in various forms. Modification examples 1 to 3 are modification examples of the first embodiment. In addition, Modification examples 4 to 9 are modification examples of the second embodiment. A part (a) of FIG. 3 illustrates a scintillator panel 1A according to Modification Example 1. The scintillator panel 1A according to Modification Example 1 has a substrate 2A, a resin reflective layer 5A, the barrier layer 3, the scintillator layer 4, and the protective film 6. The constituent material of the substrate 2A is not particularly limited. The substrate 2A may be constituted of a metal material and a carbon material and may be constituted of a glass material and/or a resin material. Examples of a metal material include aluminum and stainless steel (SUS). Examples of a carbon material include amorphous carbon, carbon fiber reinforced plastic (CFRP). Examples of a resin material include polyethylene terephthalate (PET), polyethylene naphthalate (PEN), and polyimide (PI). The resin reflective layer 5A reflects light generated in the scintillator layer 4. The resin reflective layer 5A may be constituted of a mixed material of a white pigment and a binder resin, for example. Examples of the white pigment include alumina, titanium oxide, yttrium oxide, and zirconium oxide. The protective film 6 may be constituted of polyparaxylylene. The protective film 6 may be formed by a chemical vapor deposition method (CVD method). A part (b) of FIG. 3 illustrates a scintillator panel 1B according to Modification Example 2. The scintillator panel 1B according to Modification Example 2 has the substrate 2, an inorganic reflective layer 8, the resin protective layer 5, the barrier layer 3, the scintillator layer 4, and the protective film 6. That is, the scintillator panel 1B according to Modification Example 2 is realized by adding the inorganic reflective layer 8 to the scintillator panel 1 according to the first embodiment. The inorganic reflective layer 8 is formed between the substrate 2 and the resin protective layer 5. Specifically, the inorganic reflective layer 8 has a reflective layer front surface 8a, a reflective layer rear surface 8b, and a reflective layer side surface 8c. The reflective layer front surface 8a faces the protective layer rear surface 5b. The reflective layer rear surface 8b faces the substrate front surface 2a. The inorganic reflective layer 8 may be constituted of a metal material, for example. Examples of a metal material include aluminum and silver. The inorganic reflective layer 8 may be a dielectric multilayer. A dielectric multilayer is a laminated film of silicon oxide (SiO2) and titanium oxide (TiO2). Examples of a resin protective layer 5 include polyimide (PI) and polyparaxylylene. According to the scintillator panel 1B, in the inorganic reflective layer 8, light generated in the scintillator layer 4 can be reflected. The resin protective layer 5 is formed between the scintillator layer 4 and the inorganic reflective layer 8. The resin protective layer 5 hinders the inorganic reflective layer 8 constituted of a metal material from directly coming into contact with the scintillator layer 4. Accordingly, occurrence of corrosion of the inorganic reflective layer 8 due to direct contact of the inorganic reflective layer 8 with the scintillator layer 4 can be curbed. When the inorganic reflective layer 8 is constituted of a dielectric multilayer, occurrence of corrosion of a metal substrate when there is a pinhole in a dielectric multilayer can be curbed by providing the resin protective layer 5. A part (a) of FIG. 4 illustrates a scintillator panel 1C according to Modification Example 3. The scintillator panel 1C has a substrate 2B, the resin reflective layer 5A, the barrier layer 3, the scintillator layer 4, and a resin film 9. The substrate 2B is constituted of a thin glass material (for example, having a thickness of 150 μm or smaller) or a carbon material such as CFRP. The substrate 2B has properties of being likely to warp. In the scintillator panel 1C, the resin film 9 is formed in order to curb warpage of the substrate 2B which may be generated at the time of forming a scintillator layer. Specifically, the resin film 9 is formed on the entire surface of the substrate rear surface 2b of the substrate 2B. The resin film 9 is a sheet member constituted of a resin material. In the resin film 9, a sheet member may be bonded to the substrate rear surface 2b. The resin film 9 may be formed through drying after being coated with a resin material. According to the resin film 9, occurrence of warpage of the substrate 2B can be curbed. A part (b) of FIG. 4 illustrates a radiation detector 10A according to Modification Example 4. The radiation detector 10A has a sealing portion 12A which differs from the radiation detector 10 according to the second embodiment. The constitutions of the barrier layer 3, the scintillator layer 4, the resin protective layer 5, and the sensor panel 11 are similar to those in the radiation detector 10 according to the second embodiment. The sealing portion 12A has the sealing plate 14 and a sealing frame 13A. The sealing frame 13A further has an inner sealing frame 17 and an outer sealing frame 18. The sealing frame 13 has a dual structure. The inner sealing frame 17 may be constituted of a resin material, for example. The outer sealing frame 18 may be constituted of a coating layer formed of an inorganic material and/or an inorganic solid material such as a glass rod, for example. According to this constitution, the scintillator layer 4 can be preferably protected from moisture. A part (a) of FIG. 5 illustrates a radiation detector 10B according to Modification Example 5. The radiation detector 10B differs from the radiation detector 10 according to the second embodiment in having no sealing portion 12 and having a protective film 6A, in place of the sealing portion 12. The constitutions of the barrier layer 3, the scintillator layer 4, and the sensor panel 11 are similar to those in the radiation detector 10 according to the second embodiment. The protective film 6A covers the protective layer rear surface 5b, the barrier layer side surface 3c, the scintillator layer side surface 4c, and the scintillator layer rear surface 4b. According to this constitution, the protective film 6A can protect a scintillator layer 4A from moisture. The protective film 6A may be selected from materials similar to those of the protective film 6. A part (b) of FIG. 5 illustrates a radiation detector 10C according to Modification Example 6. The radiation detector 10C is realized by further adding a sealing frame 13B to the radiation detector 10B according to Modification Example 5. Accordingly, the scintillator layer 4, the barrier layer 3, the resin protective layer 5, the sensor panel 11, and the protective film 6A are similar to those in the radiation detector 10B according to Modification Example 5. The sealing frame 13B blocks a joining portion of the resin protective layer 5 and the protective film 6A. Accordingly, when viewed in the thickness direction in a plan view, the sealing frame 13B is forming along an outer edge of the protective film 6A. The sealing frame 13B may be constituted of a UV curable resin, for example. According to this constitution, invasion of moisture through the joining portion between the sensor panel 11 and the protective film 6A is curbed. Accordingly, the moisture resistance of the radiation detector 10C can be further enhanced. A part (a) of FIG. 6 illustrates a radiation detector 10D according to Modification Example 7. The radiation detector 10D differs from the radiation detector 10 according to the second embodiment in having no sealing portion 12 of the radiation detector 10 according to the second embodiment and having a sealing sheet 12B, in place of the sealing portion 12. The constitutions of the barrier layer 3, the scintillator layer 4, the resin protective layer 5, and the sensor panel 11 are similar to those in the radiation detector 10 according to the second embodiment. The sealing sheet 12B exhibits a rectangular shape, a polygonal shape, or a circular shape in a plan view in the thickness direction. The sealing sheet 12B may be constituted of a metal foil, a metal sheet such as an aluminum sheet, or a barrier film, for example. The sealing sheet 12B covers the scintillator layer 4 and the barrier layer 3. Specifically, it covers the scintillator layer rear surface 4b, the scintillator layer side surface 4c, the barrier layer side surface 3c, and a portion of the protective layer rear surface 5b. In a plan view, the sealing sheet 12B is larger than the scintillator layer 4 and the barrier layer 3. An outer circumferential edge 12a of the sealing sheet 12B adheres to the panel front surface 11a using an adhesive 15. Accordingly, the sealing sheet 12B and the sensor panel 11 form an air-tight region containing the scintillator layer 4 and the barrier layer 3. Accordingly, the scintillator layer 4 can be protected from moisture. The adhesive 15 may include filler materials. The particle sizes of the filler materials are smaller than the thickness of the adhesion layer. A part (b) of FIG. 6 illustrates a radiation detector 10E according to Modification Example 8. The radiation detector 10E has a sealing frame 12C having a constitution different from that of the sealing sheet 12B according to Modification Example 7. The sealing frame 12C exhibits a box shape. The sealing frame 12C has an opening on a bottom surface. The sealing sheet 12B according to Modification Example 7 has flexibility. On the other hand, the sealing frame 12C according to Modification Example 8 maintains a predetermined shape and is hard. Accordingly, the sealing frame 12C may be constituted of a glass material, a metal material, or a carbon material, for example. The bottom surface of the sealing frame 12C adheres to the panel front surface 11a using the adhesive 15. According to this constitution, the scintillator layer 4 is disposed in an air-tight region formed by the sealing frame 12C and the sensor panel 11. Accordingly, in the radiation detector 10E, the scintillator layer 4 can be protected from moisture. Moreover, the sealing frame 12C is hard. As a result, in the radiation detector 10E, the scintillator layer 4 can be protected mechanically. A part (a) of FIG. 7 illustrates a radiation detector 10F according to Modification Example 9. The radiation detector 10F has a barrier layer 3A, the scintillator layer 4A, and a resin protective layer 5B which differ from those in the radiation detector 10 according to the second embodiment. The barrier layer 3A has the barrier layer front surface 3a, the barrier layer rear surface 3b, and the barrier layer side surface 3c. The scintillator layer 4A has the scintillator layer front surface 4a, the scintillator layer rear surface 4b, and the scintillator layer side surface 4c. The resin protective layer 5B has the protective layer front surface 5a, the protective layer rear surface 5b, and the protective layer side surface 5c. The single body constitution of the sensor panel 11 is similar to that in the radiation detector 10 according to the second embodiment. The scintillator layer 4A is formed on one side surface of the sensor panel 11 such that it protrudes from the photo-detection region S1. Specifically, first, the resin protective layer 5B is formed on the photo-detection region S1 and a peripheral region S2a on one side. Next, the barrier layer 3A is formed on the photo-detection region S1, the panel side surface 11c on one side, and the peripheral region S2a between the photo-detection region S1 and the panel side surface 11c on one side such that the resin protective layer 5B is covered. Then, the scintillator layer 4A is formed on the entire surface of the barrier layer 3A such that the barrier layer 3A is covered. The radiation detector 10F having this constitution can be preferably used as a radiation detector for mammography. In such application of the radiation detector 10F, the scintillator layer 4A is disposed such that a side formed to protrude from the photo-detection region S1 is positioned on the breast-wall side of an examinee. A part (b) of FIG. 7 illustrates a radiation detector 10G according to Modification Example 10. The radiation detector 10G according to Modification Example 10 has the substrate 2, the resin protective layer 5, the barrier layer 3, the scintillator layer 4, and the sensor panel 11. In the radiation detector 10Q the scintillator layer front surface 4a is attached to the sensor panel 11 such that it faces the panel front surface 11a. According to this constitution, an exposed region S3 in the substrate front surface 2a also faces the surrounding region S2 of the panel front surface 11a. The protective layer front surface 5a is separated from the panel front surface 11a as much as the heights of the scintillator layer 4 and the barrier layer 3. The sealing frame 13 is sandwiched between the protective layer front surface 5a and the panel front surface 11a. The sealing frame 13 and the resin protective layer 5 are fixed to each other through adhesion. Similarly, the sealing frame 13 and the sensor panel 11 are fixed to each other through adhesion (when the sealing frame 13 has adhesive properties, they adhere to each other through bonding, and when it has non-adhesive properties, an adhesive is provided in the interface). According to this constitution, the substrate 2 having the resin protective layer 5 can exhibit a function as a growth substrate for the barrier layer 3 and the scintillator layer 4, and a function as a sealing plate in the radiation detector 10G. Accordingly, the number of components constituting the radiation detector 10G can be reduced. In the experimental example, effects of improvement in moisture resistance exhibited by the barrier layer, have been confirmed. The moisture resistance stated in the present experimental example denotes a relationship between a time being exposed to an environment having predetermined humidity and a degree of change in resolution (CTF) indicated by the scintillator panel. That is, high moisture resistance denotes that the degree of deterioration in resolution indicated by the scintillator panel is low even when it is exposed to a humidity environment for a long time. On the contrary, low moisture resistance denotes that the degree of deterioration in resolution indicated by the scintillator panel is high when it is exposed to a humidity environment for a long time. In the experimental example, first, three test bodies (scintillator panels) were prepared. Each of the test bodies had a scintillator layer and a substrate. Each of the scintillator layers included CsI as a main component, and the thickness thereof was 600 micrometers. Then, first and second test bodies had a barrier layer including Til as a main component between the substrate and the scintillator layer. On the other hand, a third test body had no barrier layer. The third test body was a comparative example in which a scintillator layer was formed directly on a substrate. The substrate of the first test body was an organic substrate including an organic material as a main component. The first test body corresponds to a scintillator panel according to a reference example. In the second test body, a resin protective film including an organic material as a main component was formed on an aluminum substrate. The second test body corresponds to the scintillator panel according to the first embodiment. The substrate of the third test body was the same as the substrate of the second test body. The constitutions of the first to third test bodies are as follows. First test body: a substrate made of an organic material, a barrier layer, and a scintillator layer. Second test body: a substrate having an organic layer, a barrier layer, and a scintillator layer. Third test body: a substrate having an organic layer, (no barrier layer), and a scintillator layer. The resolution of each of the first to third test bodies was obtained. The resolutions were adopted as reference values. Next, the first to third test bodies were installed in an environment testing machine in which the temperature was 40° C. and the humidity was set to 90%. Next, the resolution of each of the test bodies was obtained every predetermined time elapsed from the installation time. Then, the degrees of the ratios of the resolutions obtained with lapse of every predetermined time to the resolutions (reference values) were calculated. That is, relative values with respect to the resolutions before the test bodies were installed in the environment testing machine were obtained. For example, when the relative value was 100 percent, it indicated that the resolution obtained after the predetermined time elapsed did not change with respect to the resolution before the test bodies were installed in the environment testing machine and the performance did not deteriorate. Accordingly, it indicated that as the relative value becomes smaller, characteristics of the scintillator panel deteriorate. A graph shown in FIG. 8 shows a relationship between the time being exposed to the foregoing environment (horizontal axis) and the relative value (vertical axis). The resolution of the first test body was measured after an hour, after 72 hours, and after 405 hours from the installation time. Measurement results were indicated as plots P1a, P1b, and P1c. The resolution of the second test body was measured after an hour, after 20.5 hours, after 84 hours, and after 253 hours from the installation time. Measurement results were indicated as plots P2a, P2b, P2c, and P2d. The resolution of the third test body was measured after an hour, after 24 hours, after 71 hours, and after 311 hours from the installation time. Measurement results were indicated as plots P3a, P3b, P3c, and P3d. The measurement results thereof were confirmed that performance of the third test body (plots P3a, P3b, P3c, and P3d) having no barrier layer deteriorated the most among the first to third test bodies. It was assumed that deterioration in performance occurred in the third test body because moisture percolated from the organic layer to the scintillator layer and deliquescence of the scintillator layer progressed with lapse of time due to the percolated moisture. On the other hand, regarding the first and second test bodies (plots P1a, P1b, and P1c; and plots P2a, P2b, P2c, and P2d) as well, it could be confirmed that the relative values tended to drop with the lapse of time. However, it was obvious that the degrees of drop in relative value indicated by the first and second test bodies were further curbed than the degree of drop in relative value indicated by the third test body. Accordingly, it has been found that deterioration in characteristics of a scintillator panel can be curbed by providing a barrier layer including TlI as a main component. It has been found that a barrier layer including TlI as a main component can contribute to improvement in moisture resistance of a scintillator panel. 1, 1A, 1B, 1C Scintillator panel 2, 2A, 2B Substrate 2a Substrate front surface 2b Substrate rear surface 2c Substrate side surface 3, 3A Barrier layer 3a Barrier layer front surface 3b Barrier layer rear surface 3c Barrier layer side surface 4, 4A Scintillator layer 4a Scintillator layer front surface 4b Scintillator layer rear surface 4c Scintillator layer side surface 5, 5B Resin protective layer 5A Resin reflective layer 5a Protective layer front surface 5b Protective layer rear surface 5c Protective layer side surface 6, 6A Protective film 7 Laminated body 8 Inorganic reflective layer 8a Reflective layer front surface 8b Reflective layer rear surface 8c Reflective layer side surface 9 Resin film 10, 10A, 10B, 10C, 10D, 10E, 10F, 10G Radiation detector 11 Sensor panel 11a Panel front surface 11b Panel rear surface 11c Panel side surface 12, 12A Sealing portion 12a Outer circumferential edge 12B Sealing sheet 12C Sealing frame 13, 13A, 13B Sealing frame 13a Frame front surface 13b Frame rear surface 13c Frame wall portion 14 Sealing plate 14a Plate front surface 14b Plate rear surface 14c Plate side surface 15 Adhesive 16 Photoelectric conversion element 17 Inner sealing frame 18 Outer sealing frame S1 Photo-detection region S2 Surrounding region S2a Peripheral region S3 Exposed region
060841494	description	Apparatus 10 of the present invention is shown in the FIGURE and comprises the following components: Dome 12, which is an airtight incinerator furnace tolerant to high temperature and high pressure, based on the autoclave theory. Tunnel section 14, which is the equipment for removing hazardous substances contained in the exhausted gas of Dome 12, using magnesium for capturing substances other than methane and methylene. Chamber 16, which is the equipment for treating hazardous substances (methane and methylene) with naphthol solution. A tall smokestack for treating exhaust gas is not required). Agent reservoir 18, which is provided from the safety aspects, in case any hazardous substance cannot be treated in the dome 16. Chamber 20 is an open-type solvent tank used for the decomposition of hazardous substances in the incinerated ash. In the FIGURE, 22 refers to ground (in some cases, the equipment may be placed underground). Conveyor device 24 for dumping industrial waste and household garbage having lower water content. Waste is dumped into dome 12 at input section 26. Agent reservoir 28 for chemical treatment in dome 16. Smokestack 30, used as the outlet for exhaust gas after treatments generated during incineration. High-frequency magnetic-field oscillation coil 32 for melting the waste material introduced into dome 12. Magnesium filter 34 (heat-tolerant). Fine-powdered pure iron filter 36. Outlet 38 for exhausting oxide (ashes) after incineration for environmentally-safe fixation treatment; and Outlet 40 for exhausting incinerated ashes after the environmentally-safe fixation treatment with water content becomes somewhat higher because of solvent treatment. DESCRIPTION OF THE INVENTION Referring now to the FIGURE, in the first globular dome 12 of the apparatus, water contained in the waste material will evaporate and turn into hydrogen and oxygen ions by setting the temperature at 698.degree. C. or higher, subsequently reacting with carbon, sulfur, and chloride compounds, which are melted and vaporized inside the furnace, especially with metal oxides to result in various metal compounds. In addition, even transparent plastic films that cannot be melted by high-frequency waves (horizontal waves) will be incinerated by vertical waves that are generated by coil 32 along with the melting (in other words, combustion) of the substances contained in a sac or an enclosure of such transparent plastic film. However, it may be necessary to set a volume of waste to be treated for each session because plastic film swells when incinerated. On the other hand, when the interior furnace temperature reaches 750.degree. C. or higher, elements start turning into ions, and the variety of ions will increase as the temperature goes up. In this manner, when the interior temperature reaches 1000.degree. C. or higher, a thermal convection phenomenon will start with gaseous substances at an extremely high temperature in the furnace and ionized elements generated inside will begin to form secondary bonding. Most of the ions forming secondary bonding will be crystallized to deposit in the lower section mixed with the incinerated ashes. Apparatus 34, 36 is provided for removing hazardous substances using appropriate agents within the tunnel 14 between the first globular dome 12 and the second dome 16 as shown in the FIGURE. The agents to be used for removing hazardous substances and the solvents to be used for removing specific metal ashes and hazardous gas (secondarily generated in the equipment) are as follows: (Solvent 1) 2-naphthol-4-sulfonic acid salt; (Solvent 2) The derivatives of the above-mentioned naphthol; (Solvent 3) 1-naphthylamine-4-sulfonic acid; (Solvent 4) The isomers of Solvent 3; (Solvent 5-1) Single-ring terpene; (Solvent 5-2) Double-ring terpene; (Solvent 5-3) Olefin, terpene. Metals and metal oxides to be used in the melting furnace are as follows: (Solvent 6) Magnesium (in a sand grain form); (Solvent 7) Fine granulated pure iron; (Solvent 8-1) Zeolite having a rhombic crystal structure (general formula: W.sub.m Z.sub.n O.sub.2n.sH.sub.2 O, wherein W: sodium, potassium, calcium and barium; Z: aluminosilicate, Si:Al ratio is larger than 1; s: arbitrary constant); (Solvent 8-2) Complex metal oxides having a spinel structure such as, barium-titanium oxide (BaTiO.sub.3), barium-tin oxide (BaSnO.sub.3) and calcium-zirconium oxide (CaZrO.sub.3). The environmentally-safe fixation of hazardous substances by adsorption with the solvents and specific metals set forth hereinabove thus results. The reactor furnace for environmentally-safe fixation hazardous substances through crystallization by lowering temperature and by reacting with the agents in the second globular dome 16 of the apparatus is shown in the FIGURE. In other words, a device (not shown) to spray (Solvent 1) toward the center of the furnace is provided. For removal of methane and hydrogen sulfide in a gaseous form that passed through the furnace 12 regardless of high temperature and high pressure, the use of the agents and the apparatus for the removal are employed. The reason why the melting furnace 12 is globular shaped is because that high-temperature and high-pressure gas are generated during incineration; and that it is a most stable shape that will stand against vaporization pressure of water contained in the raw material to be incinerated. Using the above procedure and apparatus, hazardous substances can be not only incinerated but also removed by use of the invention 10. By using the apparatus and the specific adsorbents and agents as stated above, exhaust gas from the melting furnace 12 can be continuously neutralized and detoxified. At the same time the ashes can be treated in a safe and stable manner to leave no secondary pollution. In the embodiment shown in the FIGURE, household garbage and industrial waste are treated. Dome 12 (first furnace) and the dome 16 (second furnace) provides for dehydration and drying. The following design parameters are utilized in implementing the invention; (1) When using far infrared ray as the heat source for drying, the wavelength must be 4 um or longer; (2) a high-frequency induction furnace is used; (3) average number of calories (thermal energy) of water is 4,190 Joule; (4) when selecting the material for the interior wall as well as strength of the furnace body, the Boyle-Charles law is used; in this regards, tungsten steel plate was chosen as the preferred material for the interior wall of the furnace; (5) gas constant is R=8.3144.times.10.sup.7 ergs per .degree.C per mole , according to Avogadro's law; (6) temperature required for obtaining thermoelectrons (operational temperature) is 1000.about.1200.degree. K.; (7) Since secondary exothermal reaction in the first furnace 12 may vary depending on the type of substance to be melted, a thermometer (not shown) is placed on the wall of the first furnace 12 so as to keep temperature between 454.degree. C. and 500.degree. C. and to induce reactants into the first furnace 12, not exceeding the melting temperature of magnesium filter 34 in the tunnel 14, i.e. 656.degree. C. The process steps are as follows: Furnace 12: household garbage and industrial waste are dehydrated and dried by far-infrared heating and introduced to dumping window, or opening, 44. After dumping, power to initiate irradition of the household garbage and industrial waste to high-frequency heat treatment is turned on. Water contained in the raw material is completely decomposed and rapidly converted into secondary ions (of oxygen and hydrogen), when the heating temperature reaches around 700.degree. C. Further, these ions will react and recombine with the group of molecules and atoms generated by melting of other substances and be incinerated into ashes through exothermal reaction. In other words, due to the autoclave reaction, interior temperatures in the furnace 12 elevate rapidly, and in one example, reached up to 2000.degree. C. As a result, sulfur oxide, hydrogen, arsenic compounds, mercury compounds, phosphorus compounds, methyl group, chloride, nitrogen oxide, oxygen, carbon, carbon monoxide, and carbon dioxide are generated. Among them, part of arsenic, mercury, and phosphorus will form oxides and hydrides and deposit to become ashes. Then, arsenic compounds, phosphorus compounds, and methyl group, etc. in the remaining portion will be conveyed to the tunnel 14 and subsequently be adsorbed by the device 36 made of fine-powdered pure iron and that of red-hot magnesium that is prepared and heated by heat ventilation from the first furnace 12. Devices 32 having internal magnesium are placed in parallel along the tunnel 14, forming a multiple-layered structure. The purpose of placing the devices 34, 36 in parallel and in a multiple-layered (multiple-stacked) structure within the apparatus 14 is to give double and triple opportunities with heated gaseous components exhausted from the first melting furnace 12. In addition, it is designed to heat magnesium held inside the device 34 through the energy of heated gas. In other words, as the thermometer placed in the first furnace 12 detects temperature of about 500.degree. C., the device for exhausting the gas to the tunnel section 14 (not shown) starts to work. On the other hand, fine-powdered pure iron is packed in the last layer 36 of the multi-layered (multiple-stacked) devices, that is the one located closest to the second reactor furnace 16. This releases carbon from carbon dioxide that is reduced by red-hot magnesium and part of such carbon turns into carbon monoxide, entering into the second furnace 16; in order to remove carbon monoxide before the intrusion of carbon monoxide into the second furnace 16, fine-powdered pure iron is employed to remove carbon monoxide by forming iron carbonyl through reaction, introduced to the second furnace 16 in that state. At this point, the reason why fine-powdered pure iron 36 is used is because the volume of adsorbing carbon monoxide is more with pure iron than magnesium (Formula 1) (Note that the formulas are set forth in Appendix A attached hereto). Tunnel Section 14: Sulfur deposits as magnesium sulfide after reacting with magnesium. The substances that are generated in the first furnace 12 and induced into the tunnel section 14 react with red-hot magnesium to become the following substances: hydrogen becomes magnesium hydroxide and deposits (however, part of sulfur and hydrogen are induced to the second furnace 16); arsenic becomes arsenic magnesium and deposits; mercury becomes mercury magnesium and deposits; phosphorus becomes phosphorus magnesium and deposits; chloride becomes magnesium chloride and deposits (however, part of it may possibly be induced to the second furnace); nitrogen becomes magnesium nitride and deposits (however, part of it may possibly be introduced to the second furnace 16); carbon becomes magnesium carbide. As described above, the process of forming reactants are exemplified in (Formula 2) and (Formula 3). In other words, various gas components generated in the first melting furnace 12, excluding those having hydrocarbon group, are mostly adsorbed by the red-hot magnesium and fine-powdered pure iron. Second Furnace 16: The device for spraying (Solvent 1) into the solution is ready. The methyl group generated in the first furnace 12 fails to react with magnesium in the tunnel 14 but reacts with excessive sulfur and hydrogen to become mercaptan (thioalcohol), which is a non-toxic crystal. In the same manner, the cyan group reacts with (Solvent 1) solution and, is subsequently hydrolyzed. The details are described in copending application Ser. No. 09/138,951, filed Aug. 24, 1998, still pending, relating to a method for fixing heavy metals and soft muddy soil due to the crystallization of oxalate. Exhaust Cooling apparatus (not shown) (located outside the melting furnace): For the apparatus, (Solvent 1) solution or (Solvent 2) solution is prepared in advance, and, if any residual substance remains in gas components exhausted from the second furnace 16, captures such substances. In addition, since no smokestack typically is provided for the apparatus, exhaust gas is passed through the solution. Assuming the case of mixing hydrocarbon such as methylene and the like into the cooling equipment, the function of fixing hydrocarbon by adding terpene liquid is provided. As to the treatment of substance deposits in the tunnel section 14, they are treated in the order of (Solvent 1) solution, (Solvent 2) solution, and graphite. The details of the treatment is set forth in copending application Ser. No. 09/138,952 filed Aug. 24, 1998, now U.S. Pat. No. 5,986,161 concurrently herewith and directed to method for stable fixing heavy metals and semi-metals as well as soft muddy soil. For heating required as one of the conditions to promote the reaction mentioned above (condensation), heat from the tunnel section 14 and the second furnace 16 is used as the source. The treatment of incinerated ashes should be completely done in the airtight chamber 20. The reason is to neutralize and detoxify hazardous gases formed by introducing again into (Solvent 3), assuming that the sulfur and chloride are vaporized during the process using (Solvent 1), (Solvent 2), and graphite material. The use of graphite is described in copending application Ser. No. 09/138,952, filed Aug. 24, 1998, now U.S. Pat. No. 5,986,161 concurrently herewith). Graphite inter-layer compounds have electron donors and receptors inserted in between the layers: the electron donors are alkaline metals, alkaline-earth metals, rare-earth metals (the III group of the periodical table Se, Y, Lanthanide, etc.), and transition metals. Electron receptors are halogen, halogenated metals, metal oxides, oxygen acids, Lewis acids, etc. At the same time, these inter-layer compounds are insoluble in water, or cannot be decomposed by water. The steps set forth hereinabove provide far more complete neutralization/detoxification of hazardous substances. According to the invention, by use of the specific agents as well as the equipment 10, it is possible to prevent secondary environmental pollution. While the invention has been described with reference to its preferred embodiments, it will be understood by those skilled in the art that various changes may be made and equivalents may be substituted for elements thereof without departing from the true spirit and scope of the invention. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the invention without departing from its essential teachings.
	description	This application is a 35 U.S.C. §371 National Phase Entry Application from PCT/IT2007/000212, filed Mar. 22, 2007, and designating the United States. The disclosure of which is incorporated herein in its entirety by reference. The present invention relates to an electrically conductive textile yarn, and to a method for production of this yarn. Yarns of this type are known for example for the production of protective garments for engineers who carry out maintenance on uncovered high-voltage cables. These garments must withstand without damage both the mechanical stresses of the cable maintenance operations and attacks by chemicals of the washing cycles, and simultaneously they must also shield the workers against any short-circuit discharges which may be generated between two distinct phases, with differences of potential which, in some states, can reach 800 kV for AC and 600 kV for DC voltages. Although the known yarns can have shielding properties, in the prior art no yarn is yet known which combines a high level of electrical conductivity with the resilience and flexibility necessary for the production of protective garments for operations on uncovered high-voltage cables, which garments are also comfortable to wear. The garments known hitherto in the prior art are made of steel wire braided with cotton thread; however these garments have a number of disadvantages: their electrical resistance, measured between any two points, is relatively high; they must be ‘activated’ before each use by means of a strong electrical discharge so as to restore their original electrical conductivity; and finally, because of its rigidity, the steel wire gives rise to rigid and uncomfortable garments which hamper the movements of the operators. It is therefore evident that it is necessary to increase the safety margins with which the operators are required to work, thus eliminating at least partially the above-described disadvantages. In view of the state of the art described, the object of the present invention is to obtain a yarn which makes it possible to produce a protective garment for operations on uncovered high voltage cables, which yarn has a high level of electrical conductivity and is also comfortable to use, so as to eliminate at least partially the above-described disadvantages. According to the present invention this object is achieved by means of a yarn as claimed herein. In the present description, the term ‘fibre’ should be understood as indicating a strand with a reduced length, for example a cut piece of wool or cotton; the term ‘thread’ indicates a plurality of spun fibres twisted together; ‘filament’ on the other hand indicates a continuous strand with a length which is substantially greater than that of the cut woollen fibres; ‘twisted yarn’ means the product obtained by twisting together one or more filaments and/or fibres; and ‘textile yarn for garments’ means a yarn which is suitable for the production of garments. As can be seen in FIG. 1 or FIG. 3, the conductive textile yarn 1 for garments according to the present invention comprises a first element 2 which is electrically conductive and a second element 3, distinct from the first element 2, which is provided with good resistance to fire. In general the second element 3 is not conductive, or its conductivity is negligible in comparison with that of the first element 2. ‘Good resistance to fire’ is understood to mean that the second element 3 has flameproof and/or flame-retardant properties that can be taken advantage of in the textile field of clothing. In this sense, this term can be understood to exclude both mineral fibres, which cannot be woven except with great difficulty, and metallic fibres which, although resistant to extremely high temperatures, are not suitable for garments for high temperatures inasmuch as they are good conductors of heat. Advantageously, the second element 3 can also be resistant to attacks by chemicals, in order to allow the yarn to withstand repeated washing cycles without being damaged. According to a preferred embodiment, the first element 2 is a conductive filament 4, preferably comprising a core 5 covered with a layer 6 of conductive metal; advantageously, the core 5 is made of nylon, and the layer 6 of conductive metal is for example made of silver. The weight of the core 5 is negligible compared with that of the filament 4: in fact, its cross-section can be less than 10%, and for example less than 3%, of the overall cross-section of the filament 4. The second element 3 comprises a thread 7, which is advantageously obtained from fibres selected from the group comprising aramide fibres, flame-retardant fibres, flame-retardant viscose, flame-retardant cotton, and Lenzing FR®. Since the yarn 1 according to the present invention advantageously also has flameproof or flame-retardant properties and a high level of chemical resistance to solvents, according to a preferred embodiment the thread 7 is made of polyaramide, and preferably meta-polyaramide, such as Nomex®, for example. The yarn 1 can also be produced by twisting one or more electrically conductive filaments 4 together with one or more threads 7. Advantageously, the resulting yarn 1 comprises a single filament 4 and a single thread 7, but it is also possible to produce yarns 1 comprising two, three or more filaments 4, and/or two, three or more threads 7, depending on the original yarn counts and that required for the final yarn. Advantageously, the fraction of weight of the electrically conductive filament 4 compared with the weight of the yarn 1 according to the present implementation method is between 5% and 50%, and preferably between 30% and 40%, for example approximately 35%. A particularly advantageous embodiment is that obtained by twisting a silver filament 4 with a yarn count of Nm 1/100000 together with a meta-aramide polymer thread 7 with a yarn count of Nm 1/55000, in order to obtain a yarn 1 with a yarn count of Nm 1/30000 (i.e. comprising approximately 35% Ag by weight and approximately 65% meta-aramide). This embodiment makes it possible to combine the desired properties in a single yarn 1; the yarn 1 can then be used to obtain garments which are provided with both a high level of electrical conductivity and the other desired properties. This embodiment requires dual processing, since, in order to obtain the yarn 1, it is necessary to twist together a filament 4 and a thread 7 in order to obtain a twisted yarn, before being able to begin to produce any garment. These disadvantages can be overcome by means of the second preferred embodiment represented in FIG. 3. In this embodiment, the first element 2 of the fibre 1 comprises one or more electrically conductive fibres 8, whereas the second element 3 comprises one or more fibres 7, similar to the fibres 7 of the first implementation method. This implementation method makes it possible to obtain a yarn 1 consisting both of conductive fibres and fibres which have good resistance to fire, so as to increase the homogeneousness of the properties of the yarn 1 itself. Advantageously the yarn 1 thus produced is twisted. The yarn 1 according to the present implementation method can be produced by mixing together one or more flocks consisting of fibres 7 with one or more flocks consisting of electrically conductive fibres 8. Advantageously, the fraction by weight of the electrically conductive fibres 8 compared with the weight of the yarn 1 according to the present implementation method is between 5% and 50%, and preferably between 20% and 40%, for example approximately 20% or 30-35%. Once the yarn 1 has been obtained, it can be used for normal textile production, including fabrics, knitwear and non-woven fabrics. It will then be possible to produce garments, such as work clothes, overalls, gloves, footwear, etc. Obviously the person skilled in the art, for the purpose of satisfying contingent and specific needs, will be able to make numerous modifications and variations to the above-described configurations. For example, the second elements 3 can be provided with anti-static properties, anti-bacterial properties, properties of resistance to temperatures higher than 200° C. and/or with particular resistance to acids or strong bases, either by means of treatment before production of the element itself, or by means of subsequent treatment. Similarly, the core 5 of the filament 4 can in turn be a continuous strand or a thread, and/or it can be produced from a material different from nylon. Furthermore, all these variants are contained within the scope of protection of the invention as defined by the following claims.
	summary
056028881	claims	1. A method for mitigating stress corrosion cracking in an oxided metal component in a water-cooled nuclear reactor, comprising the steps of: injecting a solution of a compound containing a noble metal into the coolant water during reactor shutdown or during heat-up with recirculation pump heat only; and causing said noble metal compound to decompose under reactor radiation conditions to release atoms of said noble metal which incorporate in said oxided metal component. immersing the oxided metal component in water; adding a noble metal compound to the water; and exposing the oxided metal component in water to electromagnetic radiation to cause noble metal compound in the vicinity of the oxide and metal component to decompose to release atoms of said noble metal which incorporate in said oxided metal component. injecting a solution of a compound containing a noble metal into the coolant water when said coolant water has a temperature less than the operating temperature of the reactor; and causing said noble metal compound to decompose under reactor radiation and thermal conditions to release atoms of said noble metal which incorporate in an oxide layer on said metal component. 2. The method as defined in claim 1, wherein exposing said noble metal compound to gamma radiation induces decomposition. 3. The method as defined in claim 1, wherein said noble metal is palladium. 4. The method as defined in claim 3, wherein said compound is an organometallic compound of palladium. 5. The method as defined in claim 4, wherein said organometallic compound is palladium acetylacetonate. 6. The method as defined in claim 1, wherein said metal component is made of stainless steel. 7. The method as defined in claim 1, wherein said metal component is made of nickel-based alloy. 8. A method of doping an oxided metal component with a noble metal, comprising the steps of: 9. The method as defined in claim 8, wherein said electromagnetic radiation is gamma radiation. 10. The method as defined in claim 8, wherein said electromagnetic radiation is ultraviolet radiation. 11. The method as defined in claim 8, wherein said noble metal is palladium. 12. The method as defined in claim 11, wherein said compound is an organometallic compound of palladium. 13. The method as defined in claim 12, wherein said organometallic compound is palladium acetylacetonate. 14. The method as defined in claim 8, wherein said metal component is made of stainless steel. 15. The method as defined in claim 8, wherein said metal component is made of nickel-based alloy. 16. The method as defined in claim 8, wherein said metal component is a component immersed in the coolant water of a nuclear reactor and said noble metal compound is added to said coolant water. 17. A method for mitigating stress corrosion cracking in a metal component in a water-cooled nuclear reactor, comprising the steps of: 18. The method as defined in claim 17, wherein said noble metal is palladium. 19. The method as defined in claim 17, wherein said noble metal compound is palladium acetylacetonate. 20. The method as defined in claim 17, wherein said metal component is made of stainless steel.
	claims	1. A confinement system comprising:an enclosure;an open-field magnetic system comprising one or more superconducting internal magnetic coils suspended within the enclosure and co-axial with a center axis of the enclosure, wherein the one or more superconducting internal magnetic coils each have a radius configured to balance the relative field strength between a plurality of point cusps and a plurality of ring cusps;a plurality of encapsulating magnetic coils co-axial with a center axis of the enclosure, the plurality of encapsulating magnetic coils having a larger diameter than the one or more internal magnetic coils of the open-field magnetic system, the plurality of encapsulating magnetic coils operable to form a magnetosphere around the open-field magnetic system; andone or more coil systems configured to supply the open-field magnetic system and the plurality of encapsulating magnetic coils with electrical currents to form magnetic fields for confining plasma within a magnetized sheath in the enclosure, wherein the magnetized sheath and plasma confined within the magnetized sheath encircle each of the one or more internal magnetic coils. 2. The confinement system of claim 1, wherein the plurality of encapsulating magnetic coils are operable to maintain a magnetic wall that prevents the plasma from expanding. 3. The confinement system of claim 1, wherein the plurality of encapsulating magnetic coils are located within the enclosure. 4. The confinement system of claim 1, wherein the one or more internal coils of the open-field magnetic system are toroidal in shape. 5. The confinement system of claim 1, wherein the enclosure is substantially cylindrical in shape. 6. The confinement system of claim 1, wherein the open-field magnetic system comprises a magnetic mirror system. 7. The confinement system of claim 1, wherein the open-field magnetic system comprises a cusp system, the cusp system comprising magnetic fields forming one or more magnetic field cusps. 8. A confinement system comprising:an enclosure;an open-field magnetic system comprising one or more superconducting internal magnetic coils suspended within the enclosure, wherein the one or more superconducting internal magnetic coils each have a radius configured to balance the relative field strength between a plurality of point cusps and a plurality of ring cusps;one or more encapsulating magnetic coils coaxial with the one or more internal magnetic coils of the open-field magnetic system, the one or more encapsulating magnetic coils operable to form a magnetosphere around the open-field magnetic system; andone or more coil systems configured to supply the open-field magnetic system and the one or more encapsulating magnetic coils with electrical currents to form magnetic fields for confining plasma within a magnetized sheath in the enclosure, wherein the magnetized sheath and plasma confined within the magnetized sheath encircle each of the one or more internal magnetic coils. 9. The confinement system of claim 8, wherein the one or more encapsulating magnetic coils have a larger diameter than the one or more internal magnetic coils of the open-field magnetic system. 10. The confinement system of claim 8, wherein the one or more encapsulating magnetic coils are external to the enclosure. 11. The confinement system of claim 8, wherein the open-field magnetic system comprises a magnetic mirror system. 12. The confinement system of claim 8, wherein the open-field magnetic system comprises a cusp system, the cusp system comprising magnetic fields forming one or more magnetic field cusps. 13. A confinement system comprising:an open-field magnetic system comprising one or more superconducting internal magnetic coils suspended within an enclosure, wherein the one or more superconducting internal magnetic coils each have a radius configured to balance the relative field strength between a plurality of point cusps and a plurality of ring cusps;one or more encapsulating magnetic coils coaxial with the one or more internal magnetic coils of the open-field magnetic system, the one or more encapsulating magnetic coils operable to form a magnetosphere around the open-field magnetic system; andone or more coil systems configured to supply the open-field magnetic system and the one or more encapsulating magnetic coils with electrical currents to form magnetic fields for confining plasma within a magnetized sheath in the enclosure, wherein the magnetized sheath and plasma confined within the magnetized sheath encircle each of the one or more internal magnetic coils. 14. The confinement system of claim 13, wherein the one or more encapsulating magnetic coils have a larger diameter than the one or more internal magnetic coils of the open-field magnetic system. 15. The confinement system of claim 13, wherein the one or more encapsulating magnetic coils are external to the enclosure. 16. The confinement system of claim 13, wherein the open-field magnetic system comprises a magnetic mirror system. 17. The confinement system of claim 13, wherein the open-field magnetic system comprises a cusp system, the cusp system comprising magnetic fields forming one or more magnetic field cusps. 18. The confinement system of claim 1, further comprising:a center magnetic coil located proximate to a midpoint of the enclosure; andtwo mirror magnetic coils comprising first and second mirror magnetic coils disposed on opposite sides of the center magnetic coil. 19. The confinement system of claim 1, wherein the one or more coil systems comprise:a center coil system configured to supply first electrical currents flowing in a first direction through a center magnetic coil;an internal coil system configured to supply second electrical currents flowing in a second direction through each of the one or more internal magnetic coils;an encapsulating coil system configured to supply third electrical currents flowing in the first direction through each of the plurality of encapsulating magnetic coils; anda mirror coil system configured to supply fourth electrical currents flowing in the first direction through each of two mirror magnetic coils. 20. The confinement system of claim 1, wherein each of the one or more internal magnetic coils comprise at least a first shielding surrounding the internal magnetic coil and each of the one more internal magnetic coils is suspended within the enclosure by at least one support. 21. The confinement system of claim 8, further comprising:a center magnetic coil located proximate to a midpoint of the enclosure; andtwo mirror magnetic coils comprising first and second mirror magnetic coils disposed on opposite sides of the center magnetic coil. 22. The confinement system of claim 8, wherein the one or more coil systems comprise:a center coil system configured to supply first electrical currents flowing in a first direction through a center magnetic coil;an internal coil system configured to supply second electrical currents flowing in a second direction through each of the one or more internal magnetic coils;an encapsulating coil system configured to supply third electrical currents flowing in the first direction through each of the one or more of encapsulating magnetic coils; anda mirror coil system configured to supply fourth electrical currents flowing in the first direction through each of two mirror magnetic coils. 23. The confinement system of claim 8, wherein each of the one or more internal magnetic coils comprise at least a first shielding surrounding the internal magnetic coil and each of the one more internal magnetic coils is suspended within the enclosure by at least one support. 24. The confinement system of claim 13, further comprising:a center magnetic coil located proximate to a midpoint of the enclosure; andtwo mirror magnetic coils comprising first and second mirror magnetic coils disposed on opposite sides of the center magnetic coil. 25. The confinement system of claim 13, wherein the one or more coil systems comprise:a center coil system configured to supply first electrical currents flowing in a first direction through a center magnetic coil;an internal coil system configured to supply second electrical currents flowing in a second direction through each of the one or more internal magnetic coils;an encapsulating coil system configured to supply third electrical currents flowing in the first direction through each of the one or more of encapsulating magnetic coils; anda mirror coil system configured to supply fourth electrical currents flowing in the first direction through each of two mirror magnetic coils. 26. The confinement system of claim 13, wherein each of the one or more internal magnetic coils comprise at least a first shielding surrounding the internal magnetic coil and each of the one more internal magnetic coils is suspended within the enclosure by at least one support.
	claims	1. A contact force evaluation method for evaluating a contact force against a supporting member of a tube bundle positioned in a fluid and supported by the supporting member, the method comprising:a contact force setting step of setting a contact force of the tube bundle;a probability density function calculation step of calculating a probability density function of a reaction force received by the supporting member from the tube bundle in response to a predetermined input, using a vibration analysis model of the tube bundle and the supporting member;a probability calculation step of calculating a probability that a reaction force equal to or higher than theset contact force occurs, based on the calculated probability density function; andan evaluation step of evaluating the set contact force, based on the calculated probability,wherein the tube bundle includes at least one tube array extending in same plane,wherein the supporting member includes at least a pair of anti-vibration members disposed on both sides of the tube array and extending along the plane so that the tube array is interposed therebetween, andwherein evaluation is performed for a contact load force which needs to be applied between the anti-vibration members and the tube array so as to suppress self-excited vibration of the tube bundle, which is supported by friction with the anti-vibration members, in a direction along the plane, against an excitation force of the fluid flowing through the tube bundle. 2. The contact force evaluation method according to claim 1,wherein the probability density function calculation step includes: performing a time history response analysis on the vibration analysis model using the predetermined input to obtain a time history response; calculating an average value and a standard deviation of the reaction force, based on the time history response; and calculating the probability density function as a normal distribution defined by the average value and the standard deviation. 3. The contact force evaluation method according to claim 1,wherein the probability density function is calculated by synthesizing a first probability density function and a second probability density function, the first probability density function corresponding to a first reaction force received by the supporting member in a tangential direction from a contact surface between the supporting member and the tube bundle, the second probability density function corresponding to a second reaction force received by the supporting member in a normal direction from the contact surface. 4. The contact force evaluation method according to claim 1,wherein the evaluation step including: estimating a wear amount which occurs between the tube bundle and the supporting member during a predetermined period, based on the calculated probability; and evaluating theset contact force, based on the estimated wear amount. 5. The contact force evaluation method according to claim 4, further comprising a power calculation step of calculating a power based on a work amount which occurs when a minute displacement is caused between the tube bundle and the supporting member due to the set contact force,wherein the evaluating step includes: estimating a slippage occurrence time caused during the predetermined period, based on the probability; and multiplying the power by the slippage occurrence time to estimate the wear amount. 6. The contact force evaluation method according to claim 4, further comprising an equivalent stiffness calculation step of calculating an equivalent stiffness of the tube bundle,wherein the evaluating step includes; calculating a decrease in the contact force, based on the wear amount and the equivalent stiffness; and evaluating the set contact force, based on whether the calculated decrease is equal to or lower than a threshold. 7. The contact force evaluation method according to claim 6,wherein the equivalent stiffness calculation step includes applying a finite-element method to the tube bundle to calculate the equivalent stiffness. 8. The contact force evaluation method according to claim 1,wherein the at least one tube array is composed of a plurality of U-shaped tubes, the U-shaped tubes extending in same plane, having a common curvature center, and having bent parts with different curvature radii from each other. 9. The contact force evaluation method according to claim 1,wherein the tube bundle is a heat-transfer tube bundle of a steam generator of a PWR nuclear power generating plant.
	abstract	Techniques for plasma injection for space charge neutralization of an ion beam are disclosed. In one particular exemplary embodiment, the techniques may be realized as a plasma injection system for space charge neutralization of an ion beam. The plasma injection system may comprise a first array of magnets and a second array of magnets positioned along at least a portion of an ion beam path, the first array being on a first side of the ion beam path and the second array being on a second side of the ion beam path, the first side opposing the second side. At least two adjacent magnets in the first array of magnets may have opposite polarity. The plasma injection system may also comprise a plasma source configured to generate a plasma in a region associated with a portion of the ion beam path by colliding at least some electrons with a gas.
	description	The application is a continuation-in-part of United States Utility Patent Application number 10/850,931 filed May 22, 2004 now abandoned in the name of the same inventor, Stuart McCord, and entitled LEAD FREE BARIUM SULFATE COMPOSITE, and claims the priority and benefit of that earlier application and all related applications, the entire disclosures of which are incorporated herein by this reference. This invention relates to generally to X-ray and Ion beam electrical insulators and particularly to polymer-metal-precursor composite insulators in which the metal-precursor component is barium sulfate. X-ray and gamma ray sources are presently being used in a wide array of medical and industrial machinery, and the breadth of such use expands from year to year. Consumer tend to notice medical and dental X-ray machines, but in addition to these applications there are baggage screening machines, CAT scan machines, non-destructive industrial inspection machinery and ion implantation machines used in the manufacture of silicon wafer computer chips. All require that high voltage generated within the device be contained, and furthermore that radiation be contained and directed. In particular, the ion implantation machinery increased in the 1980's and 1990's with the silicon chip boom. In the past, lead itself or lead-polymer composites were used to make electrical insulator items. But there are numerous problems with the use of lead. One problem with lead is that it is toxic and thus subject to increasingly stringent legal controls. Another issue is that lead may not have the mechanical or electrical properties desired for a given application. Lead has been used in various forms in wide range of applications: machined, as a solid casting, as a solid encased within a matrix such as a polymer matrix, or as a filler. As a filler, it may be lead particles, tribasic lead-sulfate or lead-oxide particles or particles of a specified shape or size, or as a mixture with other materials such as tin. Tungsten shielding, or polymer-tungsten shielding has also been used. Examples of all of these methods may be found in the prior art. In general, polymer-metal composites are materials having a polymer matrix containing particles of a metal compound intermixed therein. The polymer may advantageously have plastic properties allowing for ease of manufacture, but a wide variety of polymers are known for use in such composites. In the prior art, lead has been a particularly favored material for its density and ease of working. Tungsten has been favored more recently, despite cost concerns. Three characteristics in particular which make such materials desirable are electrical non-conductivity, radiological shielding ability, and high density. There is a growing list of applications for which polymer-metal composite materials are either required or advantageous. Ion implantation machine source insulators, X-ray tube insulation, radioisotope housings, other castings and housings could benefit from the properties of polymer-metal composite materials. In the case of typical high voltage insulators for ion implantation machinery, a thick walled generally round or cylindrical part is created out of lead or polymer-lead-oxide ranging from an inch to several feet or more in long dimension and weighing anywhere up to 500 pounds. Wall thickness may range from ½ inch to several inches. Such parts must resist high voltages, shield against x-ray or gamma ray emission and hold a high vacuum state when connected to the vacuum chamber. High voltage X-ray shielding for X-ray tube insulators is generally thinner (often 0.070 inch thickness), generally smaller, and of different shape, having an aperture for the X-ray beam, but once again must offer high voltage insulation and radiation protection. The lead in such devices obviously presents an environmental challenge to manufacture, use and disposal. In the processing of lead precursor filled plastics known in the art, specialized facilities, handling procedures, training and safety equipment must be used to protect the employees from the lead precursor they handle. Lead-based dust is a particular concern, being airborne and inhalable. Such dust may be generated during mixing, molding, deflashing, machining and finishing of final products such as insulators or shields, to say nothing of earlier stages of mining, smelting and refining of lead and the final disposal of the used product at the end of its useful life. Even during the life span of the product, it is illegal to sand, machine, alter or use the product in any way that will generate dust. All such processes must be carried out at special lead handling sites, and all waste dust from any of these processes must be collected in accordance with OSHA regulations and transported to hazardous waste land fills in accordance with OSHA and DES guidelines. Various radio-opaque agents are known which are used for diverse applications. Importantly, however, certain families of compounds are disfavored as having many of the same issues as lead and lead oxides. For example, the barium family of compounds are almost without exception subject to regulation due to their toxic nature. It is not previously known to use such barium family compounds in amounts greater than 10% by volume, since the structures in which they are emplaced are radio-opaque, not radiation barriers. Internalized by law into the manufacturing process, such safety issues dramatically increase the cost of such products, which in turn increases other medical or industrial costs. One recent invention to deal with this issue is TUNGSTEN-PRECURSOR COMPOSITE, for which application Ser. No. 10/095,350 filed Mar. 9, 2002 in the name of the same inventor, Stuart J. McCord was filed and has been allowed. This invention addresses material and cost concerns of tungsten shielding by proposing the use of tungsten precursor materials which testing reveals to have favorable properties. However, an entire range of desirable properties is not attainable with a single family of compounds, and so additional compounds may be desirable in order to expand the range of properties which may be attained in a lead-free shield device. Cost, of course, is one issue. Availability is another, as are actual material properties. During prosecution of that patent, U.S. Pat. No. 5,548,125 issued to Sandback (RADIATION PROTECTIVE GLOVE) and U.S. Pat. No. 4,957,943 issued to McAllister et al (PARTICLE-FILLED MICROPOROUS MATERIALS) were cited by the examiner prior to allowance. However, the glove patent, for example, teaches a flexible material most likely to be extruded. Other prior art cited includes U.S. Pat. No. 3,473,028 issued to Curry for X-RAY TUBE HOUSEING CONSISTING OF A DIELECTRIC MATERIAL WITH AN ELECTRICALLY CONDUCTIVE LINER, issued Oct. 14, 1969. The device disclosed is neither annular nor composed of truncated cone shapes. Much more importantly, it teaches towards use of a specific dielectric material and thus teaches away from the material of the invention, and for that reason may not be combined with prior art showing the materials of the present invention. U.S. Pat. No. 5,443,775 to Brannon on Aug. 22, 1995 for PROCESS FOR PREPARING PIGMENTED THERMOPLASTIC POLYMER COMPOSITIONS AND LOW SHRINKING THERMOSETTING RESIN MOLDING COMPOSITION is directed towards making of desirable colors and refractive properties in polymer products and is thus not relevant prior art for the present invention. U.S. Pat. No. 4,938,233 issued to Orrison, Jr. for RADIATION SHIELD on Jul. 3, 1990 teaches a flexible radiation shield not manufacturable by casting and not having thick walls suitable for high voltage insulation. Since the device teaches flexibility, it teaches away from thick walls and thus cannot be combined with a device having useful high voltage insulation properties (i.e. having thick walls). U.S. Pat. No. 7,079,624 to Miller et al for X-RAY TUBE AND METHOD OF MANUFACTURE, granted Jul. 18, 2006, teaches a device having an entirely different configuration, and teaches away from barium sulfate in a polymer matrix. Another attempt to deal with the issue of environmental lead contamination may be found in U.S. Pat. No. 6,048,379 issued Apr. 11, 2000 to Bray et al for “HIGH DENSITY COMPOSITE MATERIAL”. This patent teaches the use of tungsten powder, a binder and a polymer to provide a composite material offering a density high enough for use as ammunition. As stated, a serious issue with the use of tungsten is that of cost. Tungsten metal is quite expensive in comparison to lead. For example, tungsten-composite materials may cost as much as 20$ per pound. U.S. Pat. Nos. 5,730,664, 5,719,352, and 5,665,808, respectively issued to Asakura, Griffin, Bilsbury all disclose metal-polymer composites for projectiles, respectively golf balls and shot pellets. Other patents from the same art (projectiles) also propose non-toxic materials. In the actual radiation shielding art itself, various patents propose polymer-metal composites of various forms. EcoMASS (a registered trademark of the PolyOne Corporation) is a combination of tungsten metal and nylon and elastomer compounds used for shielding, apparently based upon the Bray '379 patent related to ammunition and thus developed specifically in response to military/sporting needs for non-toxic ammunition. It does not teach that materials other than tungsten may be used, thus limiting the range of characteristics of the final product. For example, tungsten is electrically conductive and thus is not normally suitable for insulators. As mentioned earlier, this material also faces cost limitations. In addition, this material has manufacturing limitations in terms of thickness and size of the final item. U.S. Pat. No. 4,619,963 issued Oct. 28, 1986 to Shoji et al for “RADIATION SHIELDING COMPOSITE SHEET MATERIAL” teaches a lead-tin fiber and resin shield, as does U.S. Pat. No. 4,485,838 issued Dec. 4, 1984 to the same inventors. Obviously the lead inclusion leads to toxicity and thus regulation questions. U.S. Pat. No. 6,310,355 issued Oct. 30, 2001 to Cadwalader for “LIGHTWEIGHT RADIATION SHIELD SYSTEM” teaches a flexible matrix having a radiation attenuating material and at least one void. U.S. Pat. No. 6,166,390 issued Dec. 26, 2000 to Quapp et al for “RADIATION SHIELDING COMPOSITION” teaches a concrete composite material. U.S. Pat. No. 5,360,666 issued Nov. 1, 1994 and U.S. Pat. No. 5,190,990 issued Mar. 2, 1993 to Eichmiller for “DEVICE AND METHOD FOR SHIELDING HEALTHY TISSUE DURING RADIATION THERAPY” teach a radiation shield for the human body comprising an elastomeric material and certain mixtures (see the summary of the invention) of various metals in the form of spherical particles. Various metals might be explored for lead replacement. In such cases, it is natural enough to skip metals having families which are generally considered toxic or too expensive, and to skip those generally used in radio-opaque applications rather than radiological blocking applications. Thus, it would be natural to skip the barium family of compounds, since these are highly regulated. It would be preferable to explore the use of other materials which are non-toxic and thus considerably safer than lead or certain available alternatives. General Summary The present invention teaches a novel lead-free plastic material that may act as a replacement for lead or lead oxide filled plastics, particularly in the role of electrical insulators in radiation devices. The present invention teaches a polymer-barium sulfate composite comprising a plastic matrix having barium sulfate materials within it as “filler” at an increased percentage of the total volume. The properties of barium sulfate are favorable and unexpected for a number of reasons. The use as an electrical insulator and materials for rigid radiation shields is unexpected due to the fact that most other members of the family are toxic and thus subject to environmental regulation, thus reducing or eliminating the key reason for lead replacement in any case. It is further unexpected in that barium sulfate is normally used in “radio-opaque” applications such as medical X-ray procedures, and it not normally considered a suitable material for actual higher density electrical insulators of radiation shielding and similar applications. The new material allows a wider range of function and use when compared with previous methods using a single metal, lead, or a lead and polymer composite. The present invention further teaches the use of binders, fibers, and secondary fillers in the polymer-barium sulfate composite in order to further broaden the range of achievable desirable physical, radiological and/or electrical properties. The present invention importantly teaches casting of the device as a process of manufacture. Summary in Reference to Claims It is a first aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure comprising: a first truncated cone section and a second truncated cone section; the two truncated cone sections secured together at their respective bases by an overlap joint; an interior space defined by the two truncated cones sections; the first and second truncated cone sections having walls, the walls made of a material comprising: a) a polymer matrix and b) barium sulfate within the polymer matrix in an approximate amount of at least 10% by volume; a first emission port passing through at least one wall; a second electrical port passing through at least one walls. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure further comprising an X-ray tube disposed within the hollow body. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure, further comprising at least one oil port passing through the walls. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure wherein the polymer matrix comprises at least one member selected from the following group: epoxy, polyester, polyurethane, silicone rubber, bismaleimides, polyimides, vinylesters, urethane hybrids, polyurea elastomer, phenolics, cyanates, cellulose, flouro-polymer, ethylene inter-polymer alloy elastomer, ethylene vinyl acetate, nylon, polyetherimide, polyester elastomer, polyester sulfone, polyphenyl amide, polypropylene, polyvinylidene flouride, acrylic, homopolymers, acetates, copolymers, acrlonitrile-butadiene-stryene, flouropolymers, ionimers, polyamides, polyamide-imides, polyacrylates, polyether ketones, polyaryl-sulfones, polybenzimidazoles, polycarbonates, polybutylene, terephthalates, polyether sulfones, thermoplastic polyimides, thermoplastic polyurethanes, polyphenylene sulfides, polyethylene, polypropylene, polysulfones, polyvinylchlorides, stryrene acrylonitriles, polystyrenes, polyphenylene, ether blends, styrene maleic anhydrides, allyls, aminos, polyphenylene oxide, and combinations thereof. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure wherein the polymer matrix comprises epoxy resin is an approximate amount of 50% to 70% by volume. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure further comprising: c) a third material. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure wherein the third material comprises at least one member selected from the following group: electrically insulating materials, binders, high density materials and combinations thereof. It is another aspect, advantage, objective and embodiment of the invention to provide a high voltage insulating radiation enclosure wherein the third material comprises at least one member selected from the following group: tungsten, lead, platinum, gold, silver, tantalum, calcium carbonate, hydrated alumina, tabular alumina, silica, glass beads, glass fibers, magnesium oxide/sulfate, wollastonite, stainless steel fibers, copper, carbonyl iron, iron, molybdenum, nickel and combinations thereof. It is another aspect, advantage, objective and embodiment of the invention to provide an electrical insulator for an ion source, the insulator comprising: a generally annular body having a diameter of at least 6 inches; the body having at least one vacuum sealing surface dimensioned and configured to provide a tight seal; at least one alignment pin projecting from the vacuum sealing surface of the insulator; at least one metal insert secured to the body; the body made of a material comprising: a. a polymer matrix and b. barium sulfate within the polymer matrix in an approximate amount of at least 35% by volume. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties, the method comprising: a) mixing uncured liquid polymers with desired percentages of powdered barium sulfate; b) blending the mixture in high shear vacuum mixers for a first predetermined time; c) placing the material into a mold having a generally annular body cavity having a diameter of at least 6 inches, the body cavity having at least one vacuum sealing surface; d) placing the material into an autoclave; e) curing it at a first temperature and first pressure for a first time. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the step a) further comprises use of epoxy polymers. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties further comprising at step a) mixing powdered hydrated alumina. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the step of mixing further comprises use of a single blade mixer. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the step of placing the mixture into a mold further comprises vacuum casting the mixture in the mold. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the step of placing the mixture into a mold further comprises pouring the mixture into the mold. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the step of placing the mixture into a mold further comprises injecting the mixture into the mold. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the first temperature comprises a range from at least 70 degrees F. to 400 degrees F. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the first time comprises a range from at least two hours to 24 hours. It is another aspect, advantage, objective and embodiment of the invention to provide a method of producing a high voltage insulator having radiation shielding properties wherein the first pressure comprises at least 50 to 250 psi. The present invention teaches novel lead-free electrical insulators of a cast plastic material that may act as replacements for lead or lead oxide filled plastics, particularly in radiation device. The presently preferred embodiment and best mode presently contemplated of the invention teaches a high voltage electrical insulator for ion implanter machines and a high voltage insulator for X-ray tube enclosures, both made of a cast polymer-barium sulfate composite comprising a high density plastic matrix having barium sulfate materials within it as filler. It is not presently known to use such barium family compounds in amounts greater than 10% by volume, since the structures in which they are emplaced in prior art are flexible and radio-opaque, not cast insulators with radiation shielding properties. Barium sulfate is a white, soluble and somewhat heavy compound normally used in paper manufacture. It is also administered prior to X-ray of patients, either as a liquid or for marking of items inserted into the patient: in either case, it's radio-opaque properties are used for internal navigation and diagnosis of patient's after the relatively low radiation exposure of such patients. By teaching the use of barium sulfate, the range of materials which may be used instead of the single metal lead is increased and thus the breadth of the properties which may be achieved is increased, another benefit of the invention. In particular, when compared to lead-composites: a. Barium sulfate consists of a combination of the barium atom, a sulfur atom, and four oxygen atoms, having properties such as a high electrical resistance, an average atomic weight of approximately 233.4, a density of roughly 4.25-4.5 grams/centimeter cubed and thus the reasonably good radiation shielding properties that are partially dependent thereon. While it does not actually meet lead oxide in terms of radiation shielding ability, it can be used in applications previously having a lower percentage of lead oxide, for example, an application having a 14% (v/v) lead component could be replaced by a component having a 35% to 45% barium sulfate component. b. Barium sulfate offers commercial advantages over tungsten metal and even over lead oxide. While a tungsten-composite may cost 20$ per pound to manufacture, and even lead oxide is roughly $1.00/lb, barium sulfate is roughly $0.30/lb at current prices, thus offering a similar or lower price. In addition, handling and manufacturing costs may be lower due to differing environmental requirements. c. Barium sulfate offer environmental advantages over lead composites. While lead causes adverse consequences after ingestion, barium sulfate does not. While lead is subject to very stringent regulations as laid out in the BACKGROUND OF THE INVENTION, barium sulfate is not. d. Barium sulfate is an unexpected choice in lead replacement applications, due to the fact that barium sulfate is the only commonly available form of barium which is not itself an environmental hazard. Thus, replacing lead in a metal-composite application with barium carbonates, nitrates, oxides, etc, would appear to be pointless in terms of avoiding hazardous material regulations, as these substances are subject to such regulation. Barium sulfate itself is relatively harmless, even being used for the infamous “barium milkshake” given to patients suffering ulcers or other gastrointestinal disorders. The barium liquid coats the interior of the GI tract and thus provides contrast during an X-ray examination of the patient. The present invention may be manufactured by casting with thermosetting materials and/or thermoplastic materials. In general, higher filler loadings may be advantageously employed. The polymers, plastics and resins which may be advantageously employed in the present invention are too numerous for a complete list, however, a partial and exemplary list includes epoxy, polyester, polyurethane, silicone rubber, bismaleimides, polyimides, vinylesters, urethane hybrids, polyurea elastomer, phenolics, cyanates, cellulose, flouro-polymer, ethylene inter-polymer alloy elastomer, ethylene vinyl acetate, nylon, polyetherimide, polyester elastomer, polyester sulfone, polyphenyl amide, polypropylene, polyvinylidene flouride, acrylic, homopolymers, acetates, copolymers, acrlonitrile-butadiene-stryene, flouropolymers, ionimers, polyamides, polyamide-imides, polyacrylates, polyether ketones, polyaryl-sulfones, polybenzimidazoles, polycarbonates, polybutylene, terephthalates, polyether sulfones, thermoplastic polyimides, thermoplastic polyurethanes, polyphenylene sulfides, polyethylene, polypropylene, polysulfones, polyvinylchlorides, stryrene acrylonitriles, polystyrenes, polyphenylene, ether blends, styrene maleic anhydrides, allyls, aminos, and polyphenylene oxide. Numerous variations and equivalents are possible. The invention is not limited to a single matrix component and a single barium sulfate composite, on the contrary multiple components may be included, for example, copolymers may be used or other mixtures of matrix elements. As another example, in tailoring of the physical properties of the composition, a blend of more than one shielding compound (such as a blend of barium sulfate and tungsten, tungsten-precursor, lead compounds, etc) may be used. In addition, the invention supports addition to the mixture of secondary fillers, binders, fibers and other components. As examples, additional electrically insulating materials, strengthening materials, materials to provide a uniform composition or bind other components, and/or density increasing materials may be used. A more specific list of examples includes such materials as tungsten metal, calcium carbonate, hydrated alumina, tabular alumina, silica, glass beads, glass fibers, magnesium oxide, wollastonite, stainless steel fibers, copper, carbonyl iron, steel, iron, molybdenum, and/or nickel. In addition, the composite material of the present invention is susceptible to a wide range of processing methods both for creation of the material and creation of items incorporating the material. In addition to casting, other techniques including molding, aggregation, machining, liquid resin casting, transfer molding, injection molding, compression molding, extrusion, pultrusion, centrifugal molding, calerending, filament winding, and other methods of handling are possible. Additionally, the composite of the invention may advantageously be worked with known equipment such as molds and machine tools, thus avoiding costs associated with re-equipping production facilities. Furthermore, since the material contains no lead, significant cost and time savings may be realized and burdensome regulations regarding lead may be properly avoided during these processes. In theory, the material may be substituted for lead oxide shielding on a basis of approximately 3.5 to 1. Thus, for typical lead oxide shielding of 0.070 inches thickness, a replacement may be manufactured at a ratio of 3.5 to 1 in thickness. In the case of liquid resin casting, this increased thickness further allows easier molding. A first formulation and embodiment of the invention was derived from barium sulfate, epoxy resin and hydrated alumina. The formulation comprised 57% by volume of an epoxy resin (438 Novolac/HHPA curative, a trademark and product of the Dow Corporation), 35% barium sulfate (catalog no. RS-22BS-35) and 8% hydrated alumina. 12 inch square plates of 0.25 inch thickness were vacuum cast and examined. Test panels were machined from the plates. The test item was compared to an equivalent lead-epoxy plate with a 14% vol/vol percentage. The cast plate was of good quality and very producible. Machined panels were of good quality, strength and durability. Material density was 0.085 lb/cubic inch, equivalent. Electrical testing showed the material to be a good insulator: Dielectric strength was 300 volts/mil per D-149, Arc resistance was 130 seconds per D-150. Shielding effectiveness was equivalent to lead oxide composite items. Despite being a barium compound, the material is non-toxic, thus despite expectations, it may be used in lead replacement roles without excessive environmental regulation. The dielectric strength was equal to the 14% lead item (300 volts/mil in both cases), and the arc resistance was approximately double that of the lead test item. This is an important factor in calculating MTBF for items made with the materials, as one source of failures is failure under arc, leading to carbon paths on the surface. Since the carbon paths are conductive, the item is rendered quickly unusable and the equipment in which it is used (micro-chip production, for example) must be shut down, interrupting manufacturing, therapy, etc. A second test item was produced, using a second formulation and embodiment of the invention derived from barium sulfate and epoxy resin. The formulation comprised 60% by volume of an epoxy resin (438 Novolac/HHPA curative, a trademark and product of the Dow Corporation) and 40% barium sulfate. 12 inch square plates of 0.25 inch thickness were vacuum cast and examined. Test panels were machined from the plates. The cast plate was of good quality and very producible. Machined panels were of good quality, strength and durability. Electrical testing showed the material to be a good insulator. Material density was 0.093 lb/cubic inch, equivalent. Shielding effectiveness was equivalent to lead oxide composite items. In summary of the test results, it can be seen that for applications requiring high resistivity and high arc resistance, barium sulfate composites may be advantageously used to achieve the desired properties. While the two tests both utilized epoxy resin, the present invention is not so limited, neither to the specific epoxy resin used nor to epoxy resin in general. Applicant reiterates that the examples presented are only examples: further development will produce numerous other materials with a wide range of characteristics, components, and methods of production. Two examples of an application of the composite are presented below, that of a ion implantation device source insulator, and a high voltage insulating X-ray box, though the invention is not so limited. It can also be seen that for applications requiring high shielding ability (such as X-ray source shielding in the medical field) the invention may be formulated to provide a shielding ability sufficient for lead replacement. Without undue experimentation higher density formulations may be produced on demand by mixing additional secondary fillers into the composition. While use of lead would under some circumstances be self-defeating, lead, tungsten, platinum, gold, iridium, silver, tantalum, and similar materials may be used. Alternatively, the barium sulfate volumetric percentage may be increased by use of injection molding, compression molding or transfer molding as permitted by materials handling techniques. As demonstrated by the example using hydrated alumina, other properties such as electrical resistivity/conductivity, workability, ductility, density, and so on may also be adjusted by use of secondary fillers, binders, and other agents in the composition. Thus it is apparent that a wide variety of products may be produced, as the characteristics of the barium sulfate composite of the present invention may be tailored depending upon the desired end characteristics. In addition, the environmental contamination engendered by the product is of a different order of magnitude than that produced by products containing lead. An exemplary list of embodiments which may advantageously be produced using the material of the present invention includes X-ray tube insulators, apertures and enclosures, X-ray tube high-voltage insulators and enclosures, X-ray tube high voltage apertures, X-ray tube high voltage encapsulation devices, high voltage insulating radioactive shielding containers and other medical X-ray and gamma ray housings. Industrially, an exemplary list of embodiments in which the composition of the invention may advantageously be incorporated include ion source insulators for ion implantation machinery and other devices for insulating, isolating, directing or shielding any radiation producing device. As stated, these lists are exemplary only and embodiments of the invention may be utilized within the art field of radiation shielding in a broad range of equivalent ways. FIG. 1 is a perspective view of an embodiment of an ion source electrical insulator according to the present invention. Ion source insulator 2 is generally annular in shape so as to allow to pass therethrough an ion implantation beam such as those used in the creation of microchip wafers. Such a device may advantageously have a desirable combination of radiation shielding ability, electrical resistivity/conductivity, physical parameters and other characteristics as are allowed by use of the polymer-barium sulfate composite of the present invention. In use, the device may be placed directly against the ion source and/or may be placed around the ion stream at later points, for example, after magnetic devices which may focus, re-direct or otherwise alter the ion beam, or in any other location in which radiation or electrical charges may need to be blocked. Vacuum sealing surfaces 10 may facilitate provision of a tight seal. Alignment pin 20, one of several possible, may be used to assure proper alignment, the number and arrangement of pins obviously allows proper alignment to be assured in as many degrees of freedom as must be restricted. Metallic inserts 30 allow attachment of the device to the overall structure of the ion implanter device, medical device, or other device to which it belongs. The inserts have internal threads (not shown) allowing easy bolting to the larger machine of which the invention will be a part or a retrofit. Such features may be produced by molding, inserts, machining, or other means suitable for use with polymer materials as are known in the art. One additional desirable quality is that these features may be created “on demand” as requested by end users of the item. Surface convolutions 40 may be used to provide additional properties such as to increase surface distance/area in order to prevent electrical arcing, to locally increase shielding or insulation, fit with other components of the overall system and so on. While the exemplary ion source insulator is quite simple, such devices may be complex, having a much greater depth, having a much greater thickness, having multiple grooves and ridges and so on. Items created using the composite of the present invention need not be annular nor even circular but may be any shape as required. The range of sizes in such insulators is quite broad: from 1 inch to 20 or more inches tall, diameters from 6 to 40 inches, wall thicknesses which might be from ½ inch thick up to 3 inches thick and weights anywhere from under 1 pound to over 500 pounds. The material of the device may be a barium sulfate composite as discussed previously. As another example, FIG. 2 teaches one example of a high voltage insulating and X-ray shielding enclosure or box. X-ray shielding insulators are typically of an extremely wide range of shapes and sizes: cylinders, three dimensional conic sections, prisms, regular and irregular solids and composite shapes. A typical “box” might be irregular, 16 inches on a side and have a weight from 1 to 30 pounds. The thickness of the walls may be even greater than that of industrial ion source insulators. The enclosure 102 shown in cross-sectional perspective in FIG. 2 is a composite of two truncated conical sections, but is an example only. It contains X-ray tube 104, having plating 106 and emitting X-ray beam 108 by means of an emission port dimensioned and configured to allow the X-ray beam to pass therethrough. Enclosure/box 102 has a number of features required to allow X-ray tube 104 to function properly. Enclosure 102 has thick walls 110 of the desired composite material: on a 3.5 to 1 replacement basis, the walls may be approximately 3.5 times as thick as a corresponding lead oxide product, but at reduced cost. Oil cooling port 120 and electrical port 130 allow oil and electrical connections to the interior of the box. Overlap joint 140 is designed to prevent radiation leakage from the joint during the case manufacture. While the exemplary ion source insulator is quite simple, such devices may be complex, having a much greater depth, having a much greater thickness, having multiple grooves and ridges and so on. Items created using the composite of the present invention need not be annular nor even circular but may be any shape as required. The range of sizes in such insulators is quite large: from 1 inch to 20 or more inches tall, diameters from 6 to 40 inches, wall thicknesses which might be from ½ inch thick up to 3 inches thick and weights anywhere from under 1 pound to over 500 pounds. High voltage insulating X-ray shielding enclosures are typically of an even wider range of shapes and sizes, cylinders, three dimensional conic sections, prisms, regular and irregular solids and composite shapes. A typical “box” might be irregular, 16 inches on a side and have a weight from 1 to 30 pounds. The thickness of the walls may be even greater than that of industrial ion source insulators. In short, regardless of shape or size of the item to be made the present invention may be adapted to any radioactive/ion/gamma ray/x-ray shielding application without undue experimentation and without departing from the scope of the invention. Formulations other than those specifically provided may be employed without departing from the scope of the invention. The method of the invention, a process for producing a high voltage insulator having radiation shielding properties, may have the following steps: TABLE IA)mixing uncured liquid epoxy polymers with desired percentages ofpowdered barium sulfate and powdered hydrated alumina.B)blending the mixture in high shear single blade vacuum mixers fora first predetermined time.C)Pouring, injecting or vacuum casting the material in a mold havinga generally annular body cavity having a diameter of at least 6 inches,the body cavity having at least one vacuum sealing surface.D)Placing the material into an autoclave.E)Curing the mold and material therein at a temperature in a range fromat least 70 degrees F. to 400 degrees F. for a period depending uponthe size, configuration and exact choice of materials, the time rangingfrom at least two hours to 24 hours, at a pressure ranging from atleast 50 to 250 psi. This is in contrast to methods of creating thin and flexible radiation barriers, which do not involve casting. This disclosure is provided to allow practice of the invention by those skilled in the art without undue experimentation, including the best mode presently contemplated and the presently preferred embodiment. Nothing in this disclosure is to be taken to limit the scope of the invention, which is susceptible to numerous alterations, equivalents and substitutions without departing from the scope and spirit of the invention. The scope of the invention is to be understood from the appended claims.
	claims	1. A computerized system for determining location by tracking a source of ionizing radiation, the system comprising:at least one first radiation detector capable of receiving ionizing radiation from the radiation source and for producing first output signals;at least one second radiation detector capable of receiving ionizing radiation from the radiation source and producing second output signals; andat least one processor configured to receive and use said first and second output signals to determine a plane in which the source resides, wherein the at least one processor is further configured to determine the plane when the source has an activity in the range of about 0.01 mCi to about 0.5 mCi. 2. The system of claim 1, wherein the at least one first radiation detector and the at one least second radiation detector are oriented to enable detection of the source when the source is connected to a medical device. 3. The system of claim 1, wherein said radiation source employs an isotope with a half life in the range of 6 to 18 months. 4. The system of claim 1, additionally comprising said radiation source capable of providing said radiation. 5. A system according to claim 1, wherein the processor is configured to determine a plane with respect to a center of mass of the source. 6. A system according to claim 1, further comprising a displacement mechanism configured to cause motion of at least one of the plurality of detectors, and wherein the at least one processor is configured to send signals to the displacement mechanism in response to radiation received, to track the radiation source. 7. A system according to claim 6, wherein the displacement mechanism is configured to track the radiation source by changing locations of detection boundaries in order to maintain the source within the detection boundaries. 8. The system of claim 1, additionally comprising:an imaging module, said imaging module being configured to provide an image signal to said at least one processor wherein said at least one processor is configured to translate the image signal into an image of a portion of a body of a subject. 9. The system of claim 8, further comprising a display device configured to display the body portion image. 10. The system of claim 9, wherein the source of radiation is connected to a medical device, and wherein said display device is configured to display the image of said portion of the body with a determined position of the medical device superimposed on said image of said portion of the body. 11. The system of claim 1, further comprising at least one third radiation detector for producing third signals, and wherein the processor is configured to determine the at least one plane using the first signals, the second signals, and the third signals. 12. The system of claim 11, further comprising an at least one fourth radiation detector for producing fourth signals, and wherein the processor is further configured to determine the at least one plane using the first signals, the second signals, the third signals and the fourth signals. 13. The system of claim 12, wherein the at least one plane is three planes. 14. The system of claim 11, wherein the at least one plane is three planes. 15. The system of claim 1, wherein the source includes a piece of radioactive metal implanted in a body of a subject. 16. The system of claim 15, wherein the piece of radioactive metal is a wire. 17. The system of claim 1, wherein each of the at least one first detector and the at least one second detector are part of at least one sensor module. 18. The system of claim 17, wherein said at least one sensor module includes at least three sensor modules. 19. The system of claim 18, further comprising at least three locomotion devices each associated with one of the at least three sensor modules, said at least three locomotion devices configured to impart translational motion to an associated sensor module in response to instructions from the at least one processor. 20. A system according to claim 18, wherein said at least three sensor modules includes at least four sensor modules, and wherein the at least one processor is configured to receive at least four output signals each defining a plane in which said radiation source resides, and to solve a resulting overdetermined set of equations to find a likely position of said radiation source, taking into account an error defined by a Euclidean distance between each plane and the position. 21. The system of claim 17, further comprising at least one locomotion device configured to impart translational motion to said at least one sensor module in response to an instruction generated by the at least one processor. 22. The system of claim 17, wherein said at least one processor receives at least two output signals each defining a plane in which said radiation source resides, and computes a linear intersection of the two planes upon which said radiation source is located. 23. The system of claim 17, wherein said at least one sensor module includes three sensor modules, and wherein the at least one processor is configured to receive at least three output signals each defining a plane in which said radiation source resides, and to compute a position of said radiation source by determining an intersection of three planes. 24. The system of claim 23, wherein said at least one processor is configured to compute said position repeatedly at predetermined intervals so that a position of said radiation source as a function of time may be plotted. 25. A method of determining a location of a radiation source, the method comprising:providing an ionizing radiation source having activity in the range of about 0.01 mCi to about 0.5 mCi;detecting from the source radiation in said range;determining based on said detected radiation a first plane in which said radiation source resides;determining based on said detected radiation at least a second plane in which said radiation source resides;locating said radiation source by calculating an intersection of said first plane and said at least a second plane. 26. The method of claim 25, wherein determining at least a second plane in which said radiation source resides includes determining at least a third plane in which said radiation source resides and wherein the method further comprises:calculating a location of intersection of said first plane, said second plane and said at least a third plane. 27. A system for determining a location associated with a source of ionizing radiation, the system comprising:at least one radiation detector configured to receive ionizing radiation from the radiation source and to produce output signals; andat least one processor configured to receive and use said output signals to determine a plane in which the source resides, wherein the at least one processor is further configured to determine the plane when the source has an activity in the range of about 0.01 mCi to about 0.5 mCi. 28. The system of claim 27, wherein the at least one radiation detector includes at least three radiation detectors, each configured to generate output signals, and wherein the at least one processor is further configured to use the output signals to identify a location within a body associated with the source. 29. The system of claim 28, wherein the at least one processor is further configured to generate instructions for aiming a therapeutic beam at the location. 30. The system of claim 28, wherein the location is an area of tissue located proximate the source. 31. The system of claim 28, wherein the at least one processor is further configured to identify relative movement between the source and the at least one radiation detector and to adjust orientations of the at least one radiation detector in response to the identified relative movement. 32. The system of claim 31, wherein the at least one processor is configured to repeatedly re-aim the at least one radiation detector toward the source in response to repeated relative movement between the source and the at least one radiation detector. 33. The system of claim 27, wherein the at least one processor is configured to cooperate with the at least one radiation detector to identify a location associated with the source when the source is a radioactive piece of metal within a body of a subject. 34. The system of claim 27, wherein the at least one processor is configured to cooperate with the at least one radiation detector, to identify a location associated with the source when the source is a radioactive piece of metal associated with a movable medical device having a portion for use within a body of a subject. 35. The system of claim 27, wherein the at least one radiation detector includes a plurality of radiation detectors, and wherein the system further includes a plurality of locomotion devices configured to receive signals from the at least one processor for re-aiming the plurality of radiation detectors toward the source when relative movement occurs between the source and the plurality of radiation detectors. 36. The system of claim 27, wherein the at least one processor is configured to identify, within a body of a subject, a location associated with the source, and wherein the at least one processor is further configured to receive information for constructing an image of the location, wherein the at least one process is further configured to output signals for displaying a location indicator superimposed on the image. 37. The system of claim 27, wherein the at least one radiation detector includes a plurality of radiation detectors configured to generate a plurality of signals, and wherein the at least one processor is configured to use the plurality of signals to identify a three dimensional location associated with the sensor. 38. The system of claim 37, wherein the at least one processor identifies the three dimensional location by calculating an intersection of at least three planes. 39. The system of claim 37, wherein the at least one processor is configured to receive information indicative of relative movement between the source and the plurality of radiation detectors, and to generate instructions for adjusting positions of the plurality of radiation detectors in response to the relative movement. 40. A system for determining a location by tracking a source of ionizing radiation, the system comprising:at least one radiation detector being configured to receive ionizing radiation from a source having a half life of between 6 and 18 months and to produce output signals; andat least one processor configured to receive and use said output signals to determine a plane in which the source resides.
	claims	1. A collimator comprising:an x-ray blocking surface comprising a one-piece plate defining an aperture formed therethrough, the aperture defining an aperture edge,wherein the aperture edge includes a first end portion including a first end of the aperture edge, a second end portion including a second end of the aperture edge, and a central portion including a center of the aperture edge, the central portion interposed between the first and second end portions, wherein the first end portion of the aperture edge corresponds to a first end portion of a detector, the second end portion of the aperture edge corresponds to a second end portion of the detector, and the central portion of the aperture edge corresponds to a central portion of the detector; andwherein a profile of the aperture edge is discontinuous at a point between the first end of the aperture edge and the center of the aperture edge. 2. The collimator of claim 1, wherein the profile of the aperture edge is discontinuous at a plurality of points between the first end of the aperture edge and the center of the aperture edge. 3. The collimator of claim 1, wherein the aperture edge comprises a linear portion and a curved portion joining the linear portion, wherein a joining point defining the location where the linear portion joins the curved portion defines a discontinuity of the aperture edge. 4. The collimator of claim 1, wherein the first end portion of the aperture edge includes a flat segment extending along a length of the aperture edge, and the central portion of the aperture edge includes a non-linear portion extending along the length of the aperture edge. 5. The collimator of claim 4, wherein the non-linear portion is configured to provide a first, substantially linear beam projection edge corresponding to the central portion of the detector, and wherein the flat segment is configured to provide a second beam projection edge that extends inwardly from the first, substantially linear beam projection in a lateral direction when the collimator is used to shape a beam for projection onto a curved detector. 6. The collimator of claim 1, wherein the aperture edge includes a plurality of differently sloped linear segments interposed between the first end of the aperture edge and the center of the aperture edge. 7. The collimator of claim 1, wherein the central portion of the aperture edge is configured to provide a first beam projection substantially conforming with a profile of the central portion of the detector, and the first end portion of the aperture edge is configured to provide a second beam projection substantially differing with a profile of the first end portion of the detector. 8. The collimator of claim 1, wherein the aperture edge is asymmetric about a center line bisecting a length of the collimator. 9. A system comprising:an x-ray source, the x-ray source providing an x-ray beam;a detector, the detector receiving a portion of the x-ray beam; anda collimator interposed between the detector and the x-ray source, the collimator comprisingan x-ray blocking surface comprising a one-piece plate defining an aperture formed therethrough, the aperture defining an aperture edge, the x-ray blocking surface configured so that the one-piece plate prevents x-ray transmission and the aperture allows x-ray transmission therethrough, wherein a projection of the beam is projected proximate to the detector;wherein the aperture edge includes a first end portion including a first end of the aperture edge, a second end portion including a second end of the aperture edge, and a central portion including a center of the aperture edge, the central portion interposed between the first and second end portions, wherein the first end portion of the aperture edge corresponds to a first end portion of the detector, the second end portion of the aperture edge corresponds to a second end portion of the detector, and the central portion of the aperture edge corresponds to a central portion of the detector; andwherein a profile of the aperture edge is discontinuous at a point between the first end of the aperture edge and the center of the aperture edge. 10. The system of claim 9, wherein the profile of the aperture edge is discontinuous at a plurality of points between the first end of the aperture edge and the center of the aperture edge. 11. The system of claim 9, wherein the aperture edge comprises a linear portion and a curved portion joining the linear portion, wherein a joining point defining the location where the linear portion joins the curved portion defines a discontinuity of the aperture edge. 12. The system of claim 9, wherein the first end portion of the aperture edge includes a flat segment extending along a length of the aperture edge, and the central portion of the aperture edge includes a non-linear portion extending along the length of the aperture edge. 13. The system of claim 12, wherein the detector is curved along a direction transverse to a central projection of the beam, wherein the non-linear portion is configured to provide a first, substantially linear beam projection edge corresponding to the central portion of the detector, and wherein the flat segment is configured to provide a second beam projection edge that extends inwardly from the first, substantially linear beam projection in a lateral direction. 14. The system of claim 9, wherein the aperture edge includes a plurality of differently sloped linear segments interposed between the first end of the aperture edge and the center of the aperture edge. 15. The system of claim 9, wherein the central portion of the aperture edge is configured to provide a first beam projection substantially conforming with a profile of the central portion of the detector, and the first end portion of the aperture edge is configured to provide a second beam projection substantially differing with a profile of the first end portion of the detector. 16. The system of claim 9, wherein the detector comprises elements in the first end portion configured for measurement of flux of the beam. 17. The system of claim 9 further comprising a processor, the processor configured to reconstruct an image using information from the detector, wherein the processor is configured to use information received from the central portion of the detector to reconstruct the image and to use information from the first end of the detector to track a focal point of the beam. 18. A system comprising:an x-ray source, the x-ray source providing an x-ray beam;a detector, the detector receiving a portion of the x-ray beam;a collimator interposed between the detector and the x-ray source, the collimator comprisingan x-ray blocking surface defining an aperture, the x-ray blocking surface comprising one or more generally flat plates defining an aperture edge of the aperture, the x-ray blocking surface configured so that the one or more generally flat plates prevent x-ray transmission and the aperture allows x-ray transmission therethrough, wherein a projection of the beam is projected proximate to the detector;wherein the aperture edge includes a first end portion including a first end of the aperture edge, a second end portion including a second end of the aperture edge, and a central portion including a center of the aperture edge, the central portion interposed between the first and second end portions, wherein the first end portion of the aperture edge corresponds to a first end portion of the detector, the second end portion of the aperture edge corresponds to a second end portion of the detector, and the central portion of the aperture edge corresponds to a central portion of the detector; andwherein the central portion of the aperture edge is configured to provide a first beam projection portion substantially conforming with a profile of the central portion of the detector, and the first end portion of the aperture edge is configured to provide a second beam projection portion substantially differing with a profile of the first end portion of the detector; anda processor configured to reconstruct an image using information provided by the detector, wherein information provided by the central portion of the detector is processed in a first manner including reconstruction of an image and information provided by the first end portion of the detector is processed in a second manner including tracking processing. 19. The system of claim 18, wherein the detector has a generally rectangular footprint and is curved along a direction transverse to a central projection of the beam, and wherein the first end portion of the aperture edge include a flat segment and the central portion of the aperture edge includes a non-linear portion. 20. The system of claim 19, wherein a joining point defining a location where the flat segment joins the non-linear portion defines a discontinuity of the aperture edge. 21. The system of claim 18, wherein the detector comprises a subgroup of elements configured to detect a flux of the second beam projection portion, the subgroup of elements positioned proximate to one or more edges of the detector.
	description	This application is a divisional of Ser. No. 12/600,774 filed on Nov. 18, 2009, which is a national stage application of International Application No. PCT/JP2008/059449 filed on May 22, 2008, which is based upon and claims the benefit of priority from Japanese Patent Application No. 2007-135302 filed on May 22, 2007; the entire contents of all of which are incorporated herein by reference. The present invention relates to a preventive maintenance/repair device and a preventive maintenance/repair method for a cylindrical structure of a cylindrical shape among reactor internal structures installed in a reactor pressure vessel. Generally, a reactor internal structure installed in a reactor pressure vessel 1 is formed of a material having an excellent corrosion resistance and a mechanical strength under a high temperature and a high pressure environment, such as an austenitic stainless steel and a nickel based alloy. However, even a reactor internal structure formed of such a material may suffer from a material deterioration which is caused by a lengthy operation under a high temperature and a high pressure environment and by an irradiation of neutron. In particular, in a portion near a welding part of the reactor internal structure, when a heat is generated upon welding, a material of the portion may be sensitized or a tensile residual stress may be generated so that a stress corrosion cracking may possibly occur. In this case, it is difficult to change the reactor internal structure for another, which poses a serious problem in terms of maintenance and administration. An overall structure of the reactor internal structure installed in the reactor pressure vessel 1 in a boiling water reactor electric-power plant (hereinafter referred to as “BWR plant”) is described with reference to FIG. 9. A shroud 2 supporting a fuel assembly is disposed inside the reactor pressure vessel 1. A jet pump 3 is disposed on a part (annulus part) between an inner wall of the reactor pressure vessel 1 and the shroud 2. In addition, disposed on a lower part of the reactor pressure vessel 1 is a control-rod drive mechanism housing 4, and disposed on an upper part of the reactor pressure vessel 1 is a core spray pipe 48 (hereinafter referred to as “CS pipe”). As shown in FIG. 10, the jet pump 3 includes: a riser pipe 8 located on a side where a coolant is taken in; a mixer nozzle 7 located above the riser pipe 8 and connected to an upper end of the riser pipe 8; and an inlet mixer 5 connected to a lower end of the mixer nozzle 7. A diffuser 6 is connected to a lower end of the inlet mixer 5. In order to support the riser pipe 8 on the inner wall of the reactor pressure vessel 1, a riser brace 9 is mounted on the inner wall of the reactor pressure vessel 1. Various maintenance/repair devices have been proposed for maintaining and repairing such a reactor internal structure. At first, there is described a case in which an outer surface of the jet pump 3 is maintained and repaired. In this case, a maintenance/repair device is firstly fitted onto a distal end of a cable or a distal end of an articulated operation pole, and the maintenance/repair device is sent into the reaction pressure vessel 1 from the upper part thereof in a hanging manner. Then, the maintenance/repair device is fixed on a reactor internal structure above the jet pump 3 or the riser brace 9. Thereafter, a target region is subjected to a maintaining and repairing operation by the maintenance/repair device. Such a maintenance/repair device is disclosed in JP11-109081A and JP2002-148385A. Next, there is described a case in which an inner surface of the jet pump 3 is maintained and repaired. In this case, the inlet mixer 5 is firstly removed. Then, the maintenance/repair device is sent into the reaction pressure vessel 1 from the upper part thereof in a hanging manner. Then, the maintenance/repair device is inserted into the jet pump 3. Thereafter, a target region is subjected to a maintaining and repairing operation by the maintenance/repair device. In an alternative method, the maintenance/repair device is inserted into the diffuser 6 from the lower part of the reaction pressure vessel 1, and a target region is subjected to a maintaining and repairing operation. Such a maintenance/repair device is disclosed in JP5-209864A and JP2003-185784A. Next, there is described a case in which the inner surface of the jet pump 3 is maintained and repaired without removing the inlet mixer 5. In this case, the maintenance/repair device is inserted into a gap formed in an opening of the mixer nozzle 7, and a target region is subjected to a maintaining and repairing operation. Such a maintenance/repair device is disclosed in JP2001-65778A, JP2002-311183A, and JP2004-251894A. However, the aforementioned maintenance/repair devices are intended to maintain and repair the diffuser 6 which is a part of the jet pump 3, and thus cannot maintain and repair the inside and the outside of the riser pipe 8. Next, there is described a case in which a wall of the shroud 2 is maintained and repaired. In this case, a maintenance/repair device is inserted into the annulus part from the upper part of the reactor pressure vessel 1. Alternatively, the maintenance/repair device is inserted from the upper part of the reactor pressure vessel 1 through an upper lattice plate 47 disposed above the shroud 2. Thereafter, a target region is subjected to a maintaining and repairing operation by the maintenance/repair device. Such maintenance/repair device are disclosed in JP Patent Nos. 3288924, 3075952, and 3069005, and JP11-174192A. However, the maintenance/repair devices are intended to maintain and repair the wall of the shroud 2, and thus cannot maintain and repair another reactor internal structure in the reactor pressure vessel 1, in particular, a cylindrical structure such as a pipe. Next, there is described a case in which a bottom part of the reactor pressure vessel 1 is maintained and repaired. Such maintenance/repair devices are disclosed in JP2002-651159A and JP Patent No. 3011583. Although these maintenance/repair devices are advantageous in their small dimensions and free movableness in water, the maintenance/repair devices cannot maintain and repair the outer surface and the inner surface of the jet pump 3, because the space of the annulus part is further narrower. Almost all the above-described maintenance/repair devices are sent into the reactor pressure vessel 1 from the upper part thereof through a cable or a wire in a hanging manner, and are brought closer to a cylindrical structure such as a pipe installed in the reactor pressure vessel 1 so as to be fixed onto the outer surface of the cylindrical structure. When the maintenance/repair device is fixed onto the inner surface of the cylindrical structure, a plurality of arms of the maintenance/repair device are expanded in a radial direction of the inner surface of the cylindrical structure. Thus, the plurality of arms are pressed onto the inner surface of the cylindrical structure, whereby the maintenance/repair device can be fixed thereon. When the aforementioned maintenance/repair device is fixed onto the outer surface of the cylindrical structure, since there are various cylindrical structures of different shapes and different dimensions, it is necessary to change the structure or the configuration of the maintenance/repair device in accordance with a target region to be maintained and repaired. Thus, the structure of the maintenance/repair device is complicated, and the size of the maintenance/repair device is further enlarged. The present invention has been made in view of the above circumstances. The object of the present invention is to provide: a preventive maintenance/repair device that is capable of precisely, circumferentially moving along an outer circumferential surface of a cylindrical structure of a cylindrical shape among reactor internal structures installed in a reactor pressure vessel, and of being securely held on the outer circumferential surface of the cylindrical structure, so as to maintain and repair the cylindrical structure; and a preventive maintenance/repair method thereof. The present invention is a preventive maintenance/repair device for use in maintaining and repairing a cylindrical structure of a cylindrical shape among reactor internal structures installed in a reactor pressure vessel, the preventive maintenance/repair device comprising: a device body; a holding mechanism connected to the device body, the holding mechanism being configured to hold the device body on an outer circumferential surface of the cylindrical structure; a traveling and driving part disposed on the device body, the traveling and driving part being configured to be circumferentially movable along the outer circumferential surface of the cylindrical structure; and a maintenance/repair mechanism disposed on the holding mechanism, the maintenance/repair mechanism being configured to maintain and repair the cylindrical structure. The present invention is the preventive maintenance/repair device wherein the holding mechanism includes: a pair of arms each having a shape along the outer circumferential surface of the cylindrical structure; guide rollers respectively disposed on distal ends of the pair of arms; arm cylinders configured to respectively drive the pair of arms; and links connected between the arms and the arm cylinders, the links being configured to transmit drives of the arm cylinders to the arms. The present invention is the preventive maintenance/repair device wherein each of the arms of the holding mechanism is separable into a proximal arm body and a distal arm end, and in order to hold the device body on an outer circumferential surface of another cylindrical structure of a different outer diameter, the arm end can be replaced with another arm end of a different length with respect to the arm body. The present invention is the preventive maintenance/repair device wherein each of the arms of the holding mechanism is separable into a proximal arm body and a distal arm end, and in order to hold the device body on an outer circumferential surface of another cylindrical structure of a different outer diameter, the arm body can be replaced with another arm body of a different length with respect to the device body. The present invention is the preventive maintenance/repair device wherein a distance sensor is disposed on an outer surface of the device body on a side opposed to a surrounding structure. The present invention is the preventive maintenance/repair device wherein a distance sensor is disposed on an outer surface of the device body on a side opposed to a surrounding structure. The present invention is the preventive maintenance/repair device wherein the maintenance/repair mechanism includes an equipment configured to maintain and repair the cylindrical structure, and an equipment cylinder configured to drive the equipment in a longitudinal direction of the cylindrical structure. The present invention is the preventive maintenance/repair device wherein the equipment of the maintenance/repair mechanism is formed of an ultrasonic flaw-detecting probe. The present invention is the preventive maintenance/repair device wherein the equipment of the maintenance/repair mechanism is a camera for a visual observation. The present invention is the preventive maintenance/repair device wherein the equipment of the maintenance/repair mechanism is a polishing jig. The present invention is the preventive maintenance/repair device wherein the device body is provided with an access device configured to bring the device body closer to the cylindrical structure so as to attach the device body to the outer circumferential surface of the cylindrical structure and to detach therefrom the device body, and an operation pole is connected to the access device, The present invention is the preventive maintenance/repair device wherein the access device includes: an access-device holding part engageable with the device body; and a rotating and driving part interposed between the operation pole and the access-device holding part, the rotating and driving part being configured to rotate the device body and the access-device holding part with respect to the operation pole. The present invention is a preventive maintenance/repair device for use in maintaining and repairing a cylindrical structure of a cylindrical shape among reactor internal structures installed in a reactor pressure vessel, the preventive maintenance/repair device comprising: a device body; a holding mechanism connected to the device body, the holding mechanism being configured to hold the device body on an outer circumferential surface of the cylindrical structure; a traveling and driving part disposed on the device body, the traveling and driving part being configured to be circumferentially movable along the outer circumferential surface of the cylindrical structure; and a maintenance/repair mechanism disposed on the device body, the maintenance/repair mechanism being configured to maintain and repair the cylindrical structure; wherein the device body is provided with a thruster driving part configured to move the device body in water. The present invention is the preventive maintenance/repair device wherein a buoyant member is located in one of the device body and the maintenance/repair mechanism. The present invention is a preventive maintenance/repair method comprising the steps of: mounting the access device and the operation pole on the device body; sending the device body into the reactor pressure vessel in a hanging manner through the operation pole and bringing the device body closer to the cylindrical structure; holding the device body on the outer circumferential surface of the cylindrical body by the holding mechanism; removing the access device from the device body, after the device body has been held on the cylindrical structure by the holding mechanism; and performing a maintenance/repair operation to the cylindrical structure by the maintenance/repair mechanism. According to the present invention, the preventive maintenance/repair device can be securely held on the outer circumferential surface of the cylindrical structure in the reactor pressure vessel. In addition, the maintenance/repair device can be circumferentially moved on the outer circumferential surface of the cylindrical structure. Thus, the maintenance/repair device can be precisely moved to a target region of the outer circumferential surface of the cylindrical structure, so that the outer circumferential surface of the cylindrical structure can be maintained and repaired. According to the present invention, the preventive maintenance/repair device can be securely held on the outer circumferential surface of the cylindrical structure in the reactor pressure vessel. In addition, the maintenance/repair device can be circumferentially moved on the outer circumferential surface of the cylindrical structure. In addition, the maintenance/repair device can be moved in water. Thus, the maintenance/repair device can be precisely moved to a target region of the outer circumferential surface of the cylindrical structure, so that the outer circumferential surface of the cylindrical structure can be maintained and repaired. Embodiments of the present invention will be described herebelow with reference to the accompanying drawings. FIGS. 1 to 6 show a preventive maintenance/repair device in a first embodiment of the present invention. FIGS. 1(a) and 1(b) are overall structural views for explaining the preventive maintenance/repair device. FIGS. 2(a) and 2(b) are views for explaining a method of determining an initial holding position at which the preventive maintenance/repair device is held. FIGS. 3(a) and 3(b) are structural views for explaining a traveling and driving part. FIGS. 4(a), 4(b), 4(c), and 4(d) are structural views for explaining a maintenance/repair mechanism. FIG. 5 is a view for explaining an access device. FIG. 6 is a view for explaining the connection of an operation pole to the access device. At first, a preventive maintenance/repair device 10 in the first embodiment of the present invention is described with reference to FIGS. 1(a) and 1(b). FIG. 1(a) is a plan view of the overall structure for explaining the preventive maintenance/repair device 10, and FIG. 1(b) is a front view of the overall structure for explaining the preventive maintenance/repair device 10. The preventive maintenance/repair device 10 in this embodiment is a device for cleaning, checking, inspecting, maintaining, and repairing an outer surface of a cylindrical structure 19 of a cylindrical shape such as a pipe among reactor internal structures installed in a reactor pressure vessel 1 in a BWR plant. In particular, target regions are welding portions of a jet pump 3, a core spray pipe 48 (hereinafter referred to as “CS pipe”), and a control-rod drive mechanism housing 4 (hereinafter referred to as “CRD housing”), which are generically referred to as a cylindrical structure 19 installed in the reactor pressure vessel 1. At first, the overall structure of the preventive maintenance/repair device 10 in this embodiment is described with reference to FIGS. 1(a) and 1(b). The preventive maintenance/repair device 10 includes: a device body 11; a pair of holding mechanisms 13a and 13b connected to the device body 11, the holding mechanism 13a and 13b being capable of holding the device body 11 on an outer circumferential surface of the cylindrical structure 19; and a traveling and driving part 20 disposed on the device body 11, the traveling and driving part 20 being capable of circumferentially moving along the outer circumferential surface of the cylindrical structure 19. Each of the holding mechanisms 13a and 13b is provided with a maintenance/repair mechanism 16 that maintains and repairs the cylindrical structure 19. The pair of holding mechanisms 13a and 13b respectively include: a pair of arms 12a and 12b each having a shape along the outer circumferential surface of the cylindrical structure 19; and guide rollers 14a and 14b disposed on distal ends of the pair of arms 12a and 12b. Respectively connected to the pair of arms 12a and 12b are arm cylinders 17a and 17b and arm cylinders 17c and 17d for driving the arms 12a and 12b pneumatically or hydraulically. The arm 12a and the arm cylinders 17a and 17b are connected by a link 18a, and the arm 12b and the arm cylinders 17c and 17d are connected by a link 8b, the links 18a and 18b being configured to transmit the driving force of the arm cylinders 17a and 17b and the driving force of the arm cylinders 17c and 17d to the arms 12a and 12b. The arms 12a and 12b can be removed from the connection portions of the links 18a and 18b, so that arm bodies 49a and 49b and arm ends 49c and 49d can be replaced with other arm bodies 49a and 49b of different lengths and other arm ends 49c and 49d of different lengths so as to correspond to a different outer diameter of another cylindrical structure 19. In addition, the arms 12a and 12b of the holding mechanisms 13a and 13b respectively have lengths corresponding to the outer diameter of the cylindrical structure 19. The arms 12a and 12b respectively include: the arm bodies 49a and 49b disposed on the side of the device body 11 (on the proximal side); and the arm ends 49c and 49d that are disposed on distal ends of the arm bodies 49a and 49b so as to be separable therefrom. Thus, in order to hold the device body 11 on an outer circumferential surface of another cylindrical structure 19 of a different outer diameter, the arm ends 49c and 49d of the arms 12a and 12b are replaceable with other arm ends 49c and 49d having lengths different from those of the former arm ends 49c and 49d so as to correspond to the different outer diameter of the other cylindrical structure 19. As shown in FIGS. 2(a) and 2(b), a distance sensor 28 is disposed on an outer surface of the device body 11 on a side opposed to a surrounding structure. The distance sensor 28 is formed of an ultrasonic distance sensor, and is capable of measuring a distance between the distance sensor 28 and a surrounding structure. FIG. 2(a) shows a state before an initial holding position at which the preventive maintenance/repair device 10 is held on the cylindrical structure 19 is not determined yet, and FIG. 2(b) shows a state after the initial holding position of the preventive maintenance/repair device 10 has been determined and the preventive maintenance/repair device 10 is held on the initial holding position. As shown in FIGS. 3(a) and 3(b), the traveling and driving part 20 disposed on the device body 11 includes a traveling wheel 21 that is driven in rotation in contact with the outer circumferential surface of the cylindrical structure 19, and a motor 26 that drives the traveling wheel 21 in rotation. FIG. 3(a) is a front view for explaining the traveling and driving part 20, and FIG. 3(b) is a bottom view of the traveling and driving part 20. A shaft 25 is connected to the traveling wheel 21, and a gear 22 is connected to the shaft 25. The gear 22 is connected to an output shaft of the motor 26 via gears 23 and 24. Further, connected to the motor 26 are cables 27a and 27b for remotely operating the motor 26. The traveling and driving part 20 is removably mounted on the device body 11. As shown in FIGS. 4(a), 4(b), 4(c), and 4(d), the maintenance/repair mechanism 16 includes an equipment 15 that maintains and repairs the cylindrical structure 19, and an equipment cylinder 30 that drives the equipment 15 in a longitudinal direction of the cylindrical body 19. FIG. 4(a) is a front view for explaining a state in which the equipment 15 of the maintenance/repair mechanism 16 is held at an upper position, FIG. 4(b) is a side view for explaining a state in which the equipment 15 of the maintenance/repair mechanism 16 is held at the upper position, FIG. 4(c) is a front view for explaining the equipment 15 of the maintenance/repair mechanism 16 is held at a lower position, and FIG. 4(d) is a side view for explaining the equipment 15 of the maintenance/repair mechanism 16 at the lower position. As shown in FIGS. 4(a), 4(b), 4(c), and 4(d), a plate 29 is removably mounted on the arm 12b of the holding mechanism 13b. The plate 29 has an equipment cylinder 30 and a slide guide 31. The equipment 15 is mounted on the slide guide 31 via a plate 32b. Fixed via a plate 32a on a surface of the equipment 15 which is opposed to the surface on which the plate 32b is disposed is an L-shaped fitting 33 to be connected to an end of a shaft of the equipment cylinder 30. As shown in FIGS. 4(a), 4(b), 4(c), and 4(d), the arm 12b is provided with a cutout 12c corresponding to the maintenance/repair mechanism 16. Although not shown, the other arm 12a, which is disposed on the side opposite to the arm 12b with respect to the cylindrical structure 19, is provided with the same cutout 12c. Thus, the maintenance/repair mechanism 16 can be mounted not only on the arm 12b but also on the arm 12a. In FIGS. 4(a), 4(b), 4(c), and 4(d), although the cutout 12c of the arm 12b faces downward, the cutout 12c may be formed to face upward. Thus, in FIGS. 4(a), 4(b), 4(c), and 4(d), the maintenance/repair mechanism 16 can be mounted on the arm 12b to face downward or upward. Similarly, in FIGS. 4(a), 4(b), 4(c), and 4(d), the maintenance/repair mechanism 16 can be mounted on the arm 12a to face downward or upward. The equipment 15 of the maintenance/repair mechanism 16 is formed of an ultrasonic flaw-detecting probe, such as a phased-array UT probe, which can ultrasonically detect a flaw such as a crack of the cylindrical structure 19 without contacting the cylindrical structure 19. Alternatively, the equipment 15 of the maintenance/repair mechanism 16 may be a camera for a visual observation. Further, the equipment 15 of the maintenance/repair mechanism 16 may be a polishing jig. As shown in FIG. 5, the device body 11 is equipped with an access device 34 that brings the device body 11 closer to the cylindrical structure 19 so as to attach the device body 11 to the outer circumferential surface of the cylindrical structure 19 and to detach therefrom the device body 11. In addition, connected to the access device 34 is an operation pole 46 via a connection part 45. The access device 34 includes: an access-device holding part 35 engageable with the device body 11; and a rotating and driving part 36 interposed between the connection part 45 on the side of the operation pole 46 and the access-device holding part 35, the rotating and driving part 36 being capable of rotating the device body 11 and the access-device holding part 35 with respect to the connection part 45 and the operation pole 46. The access-device holding part 35 includes a plurality of device-body side pins 35a to be inserted into a plurality of holes formed in the device body 11, holding pins 35b for holding the device-body side pins 35a, and holding cylinder 37 for pressing the holding pins 35b onto the device-body side pins 35a via fittings 35c. The rotating and driving part 36 includes a frame 38 rotatably connected to the access-device holding part 35 through a pin 42, and a rotational cylinder 40. A fixed side of the rotational cylinder 40 is rotatably connected to the frame 38 through a pin 39, and an end of a shaft 43 is rotatably connected to a fixed side of the holding cylinder 37 through a pin 44. The frame 38 has the connection part 45 to which the operation pole 46 can be connected. As shown in FIG. 6, the operation pole 46 is composed of an operation pole 46a and an operation pole 46b. The operation pole 46a is connected to the connection part 45 of the frame 38 (see, FIG. 5), and the operation pole 46a and the operation pole 46b are connected to each other. Next, an operation of this embodiment as structured above is described. Given herein as an example to describe the process is a case where a pipe installed vertically in the reactor pressure vessel 1 is maintained and repaired. How to fit the access device 34 on the device body 11 is described with reference to FIG. 5. At this time, the plurality of device-body side pins 35a of the access-device holding part 35 disposed on the access device 34 are firstly inserted into the plurality of holes formed in the device body 11. Then, the holding cylinders 37 of the rotating and driving part 36 are driven so as to press the holding pins 35b onto the device-body side pins 35a via the fittings 35c. Thus, the device body 11 can be held by the access device 34, without the device-body side pins 35a coming off from the holes formed in the device body 11. Next, how to fit the operation pole 46a on the connection part 45 of the frame 38 (see, FIG. 5) disposed on the access device 34 is described with reference to FIG. 6. At this time, one end of the operation pole 46a is connected to the connection part 45 of the frame 38 of the rotating and driving part 36 of the access device 34. Then, the operation pole 46b is connected to the other end of the operation pole 46a. Thereafter, the device body 11 hanging from the operation pole 46 composed of the operation poles 46a and 46b is sent into the reactor pressure vessel 1, and is brought closer to the cylindrical structure 19. At this time, as shown in FIG. 5, the rotation cylinder 40 of the rotating and driving part 36 of the access device 34 is driven so as to contract the shaft 43 of the rotation cylinder 40. Thus, the device body 11 and the access-device holding part 35 are rotated with respect to the operation pole 46a, so that the device body 11 is oriented in substantially the same direction as the longitudinal direction of the operation poles 46a and 46b. Then, as shown in FIG. 6, the preventive maintenance/repair device 10, which is held by the operation poles 46a and 46b connected to the access device 34, is sent in the hanging manner via the access device 34. After that, the preventive maintenance/repair device 10 is sent into the reactor pressure vessel 1 through the lateral side of the CS pipe 48, and is brought closer to a target region of the cylindrical structure 19. Then, as shown in FIG. 5, the rotation cylinder 40 is driven so as to extend the shaft 43 of the rotation cylinder 40. Thus, the device body 11 and the access-device holding part 35 are rotated with respect to the operation pole 46a, so that the device body 11 is oriented in a direction substantially perpendicular to the longitudinal direction of the operation poles 46a and 46b. Then, the device body 11 is held by the holding mechanism 13 on the outer circumferential surface of the cylindrical structure 19. At this time, as shown in FIG. 1(a), the traveling wheel 21 of the traveling and driving part 20 is in contact with the outer circumferential surface of the cylindrical structure 19. In addition, by driving the arm cylinder 17a, 17b, 17c, and 17d of the holding mechanisms 13a and 13b, the guide rollers 14a and 14b disposed on the distal ends of the arms 12a and 12b are pressed onto the outer circumferential surface of the cylindrical structure 19 via the links 18a and 18b and the arms 12a and 12b. Thus, the preventive maintenance/repair device 10 is held on the outer circumferential surface of the cylindrical structure 19. In a case in which another cylindrical structure 19 of a different outer diameter is maintained and repaired, the arm ends 49c and 49d of the arms 12a and 12b are previously replaced with other arm ends 49c and 49d so that the arms 12a and 12b have the lengths corresponding to the outer diameter of this cylindrical structure 19. Thus, the preventive maintenance/repair device 10 can be securely held on the cylindrical structure 19 of a given outer diameter. As described above, after the device body 11 has been held on the cylindrical structure 19 by the arms 12a and 12b of the holding mechanism 13, the access device 34 is removed from the device body 11. At this time, the holding cylinder 37 of the access-device holding part 35 is driven so as to move the holding pins 35b via the fittings 35c from the device-body side pins 35a to a side opposed to the device-body side pins 35a. Then, the access-device holding part 35 hanging from the operation poles 46a and 46b is moved upward of the reactor pressure vessel 11. Thus, the plurality of device-body side pins 35a can be drawn out from the plurality of holes formed in the device body 11. Thereafter, the access device 34 is moved further upward of the reactor pressure vessel 1 in the hanging manner by means of the operation poles 46a and 46b, and the access device 34 is withdrawn. Then, the preventive maintenance/repair device 10 is circumferentially moved along the outer circumferential surface of the cylindrical structure 19. In this case, a drive command is firstly given from a remote location to the motor 26 of the traveling and driving part 20 through the cables 27a and 27b. At this time, the motor 26 is driven, so that the rotational force of the motor 26 is transmitted to the shaft 25 via the gear 24, the gear 23, and the gear 22. The rotational force of the shaft 25 is transmitted to the traveling wheel 11 connected to the shaft 25, so that the traveling wheel 11 is drive in rotation. Thus, the preventive maintenance/repair device 10 can be circumferentially moved along the outer circumferential surface of the cylindrical structure 19. When the preventive maintenance/repair device 10 is circumferentially moved along the outer circumferential surface of the cylindrical structure 19, the guide rollers 14a and 14b disposed on the distal ends of the arms 12a and 12b of the holding mechanisms 13a and 13b are rotated in accordance with the movement of the preventive maintenance/repair device 10. Also at this time, as described above, the guide rollers 14a and 14b are pressed onto the outer circumferential surface of the cylindrical structure 19 by the arm cylinders 17a, 17b, 17c, and 17d of the holding mechanisms 13a and 13b. Thus, the preventive maintenance/repair device 10 can be smoothly, circumferentially moved along the outer circumferential surface of the cylindrical structure 19, while the preventive maintenance/repair device 10 is being held on the outer circumferential surface of the cylindrical structure 19. As a result, the equipment 15 of the maintenance/repair mechanism 16 can be smoothly moved to a desired circumferential position along the outer circumferential surface of the cylindrical structure 19. The traveling and driving part 20 can be easily mounted on and removed from the device body 11. Thus, if the traveling and driving part 20 is broken for some reason or another, the whole preventive maintenance/repair device 10 is drawn upward, and the broken traveling and driving part 20 can be replaced with another normal traveling and driving part 20, which has been prepared beforehand, for a short period of time. Then, as shown in FIGS. 2(a) and 2(b), there is determined an initial position at which the preventive maintenance/repair device 10 is held on the outer circumferential surface of the cylindrical structure 19. The reason for determining the initial position is as follows. In order to hold the preventive maintenance/repair device 10 on the cylindrical structure 19, since the space of the annulus part surrounding the cylindrical structure 19 is narrow, it is necessary to bring the preventive maintenance/repair device 10 closer to a target region of the cylindrical structure 19 from a direction in which the preventive maintenance/repair device 10 does not interfere with a structure surrounding the cylindrical structure 19. Thus, the direction in which the preventive maintenance/repair device 10 is held varies for each time, i.e., the direction in which the preventive maintenance/repair device 10 is held is not constant. During the operation, there is a possibility that the preventive maintenance/repair device 10 is not held on a target region of the cylindrical structure 19 so as to be detached from the target region for some reason or another. In this case, the equipment 15 of the maintenance/repair mechanism 16 is not appropriately opposed to the outer circumferential surface of the cylindrical structure 19. Under this state, the target region of the cylindrical structure 19 cannot be precisely maintained and repaired. It is difficult to exactly return the preventive maintenance/repair device 10 to the original target region from which the preventive maintenance/repair device 10 has been detached. Even when the preventive maintenance/repair device 10 is returned to a region near the original target region from which the preventive maintenance/repair device 10 has been detached so as to continue the maintenance/repair operation, there may remain some region that could not be maintained and operated, between the original target region from which the preventive maintenance/repair device 10 has been detached and the region to which the preventive maintenance/repair device 10 is returned. In order to avoid this situation, in this embodiment, there is determined an initial holding position at which the preventive maintenance/repair device 10 is held on the outer circumferential surface of the cylindrical structure 19. At this time, there is used the distance sensor 28 which is disposed on the outer surface of the device body 11 on a side opposed to the surrounding structure, which is shown in FIGS. 2(a) and 2(b). At first, by driving the motor 17 of the traveling and driving part 20, the preventive maintenance/repair device 10 is circumferentially moved along the outer circumferential surface of the cylindrical structure 19, and a distance between the distance sensor 28 and the surrounding structure is simultaneously measured. Then, there is obtained a position of the preventive maintenance/repair device 10 at which the measured distance is shortest. The thus obtained position is recorded, and the position is determined as an original point of the initial holding position of the preventive maintenance/repair device 10. Below the guide roller 14a or 14b of the holding mechanism 13a or 13b, there are disposed a rotational sensor (not shown) that measures an angle of a holding position of the preventive maintenance/repair device 10, and a measurement wheel (not shown) that rotates the rotational sensor. With the use of the rotational sensor, an angle of the holding position of the preventive maintenance/repair device 10 is measured. Thus, there can be obtained an angle between the aforementioned original point and the holding position of the preventive maintenance/repair device 10 at which the preventive maintenance/repair device 10 has been circumferentially moved along the outer circumferential surface of the cylindrical structure 19 from the original point. In this manner, an initial holding position of the preventive maintenance/repair device 10 is determined. Thus, even when the preventive maintenance/repair device 10 is detached from the target region of the cylindrical structure 19 during the maintaining and repairing operation, it is possible to specify the position at which the preventive maintenance/repair device 10 has been held before the preventive maintenance/repair device 10 is detached therefrom. Accordingly, the maintaining and repairing operation can be reliably performed both at the holding position in the target region from which the preventive maintenance/repair device 10 has been detached and at the holding position in the target region to which the preventive maintenance/repair device 10 is returned. Then, the equipment 15 of the maintenance/repair mechanism 16 is moved by the equipment cylinder 30 in the longitudinal direction of the cylindrical structure 19. At this time, the shaft of the equipment cylinder 30 for moving the equipment 15 in the longitudinal direction of the cylindrical structure 19 is driven so as to be expanded and contracted. Since the equipment 15 is slidably disposed in the longitudinal direction of the cylindrical structure 19 with respect to the fixed plate 29 by the slide guide 31, the expansion and contraction drive of the shaft of the equipment cylinder 30 is transmitted to the equipment 15 via the L-shaped fitting 33 and the plate 32a, so that the equipment 15 is slid in the longitudinal direction of the cylindrical structure 19. Thus, the equipment 15 can be precisely moved in the longitudinal direction of the cylindrical structure 19 toward a target region of the cylindrical structure 19. Further, the plate 29 of the maintenance/repair mechanism 16 can be fixed not only to the arm 12b but also to the arm 12a. Furthermore, the maintenance/repair mechanism 16 disposed to face downward in FIGS. 4(a), 4(b), 4(c), and 4(d) may be disposed to face upward in FIGS. 4(a), 4(b), 4(c), and 4(d). Thus, when a working space in which the maintaining and repairing operation is performed by the preventive maintenance/repair device 10 is restricted, by changing the position and the direction of the maintenance/repair mechanism 16, the maintaining and repairing operation is performed over a wide range as much as possible by effectively using the working space. Then, by driving the motor 17 of the traveling and driving part 20, the preventive maintenance/repair device 10 is circumferentially moved along the outer circumferential surface of the cylindrical structure 19. Thus, the equipment 15 can be circumferentially moved along the outer circumferential surface of the cylindrical structure 19. Accordingly, the cylindrical structure 19 can be maintained and repaired by the equipment 15 over all the outer circumferential surface of the cylindrical structure 19. When the equipment 15 of the maintenance/repair mechanism 16 shown in FIGS. 4(a), 4(b), 4(c), and 4(d) is formed of an ultrasonic flaw-detecting probe, a target region of the cylindrical structure 19 can be ultrasonically detected. Thus, whether there is a crack or not in a welding line of the cylindrical structure 19 can be checked. Alternatively, the equipment 15 of the maintenance/repair mechanism 16 is formed of a camera for a visual observation, the appearance of the outer circumferential surface of the cylindrical structure 19 can be visually checked. In the above embodiment, there has been described the maintaining and repairing operation which is performed when the reactor pressure vessel 1 is filled with water. However, the reactor pressure vessel 1 is not filled with water but with air, the equipment 15 of the maintenance/repair mechanism 16 may be replaced with another equipment so as to perform another maintaining and repairing operation. Namely, when the equipment 15 is formed of a polishing jig, the outer circumferential surface of the cylindrical structure 19 can be repaired. Alternatively, when the equipment 15 is formed of a cleaning equipment such as a brush or a water-cleaning nozzle, the outer circumferential surface of the cylindrical structure 19 can be cleaned. Alternatively, when the equipment 15 is formed of a maintenance equipment such as a water-jet peening head or a laser peening head, the cylindrical structure 19 can be maintained. Alternatively, when the equipment 15 is formed of a welding head or a grinding jig, the outer circumferential surface of the cylindrical structure 19 can be repaired. According to this embodiment, the preventive maintenance/repair device 10 can be securely held on the outer circumferential surface of the cylindrical structure 19 installed in the reactor pressure vessel 1. In addition, the preventive maintenance/repair device 10 can be circumferentially moved on the outer circumferential surface of the cylindrical structure 19. Thus, the preventive maintenance/repair device 10 can be precisely moved to a target region of the outer circumferential surface of the cylindrical structure 19, so that the outer circumferential surface of the cylindrical structure 19 can be maintained and repaired. Next, a preventive maintenance/repair device in a second embodiment of the present invention is described with reference to FIGS. 7 and 8. FIGS. 7(a) and 7(b) are overall structural views for explaining the preventive maintenance/repair device. FIGS. 8(a) and 8(b) are structural views for explaining a thruster driving part. The second embodiment shown in FIGS. 7(a) and 7(b) and FIGS. 8(a) and 8(b) is embodied by applying a thruster driving part to the first embodiment, and other structures of the second embodiment are substantially the same as those of the first embodiment shown in FIGS. 1 to 6. In this embodiment, the parts and elements identical to those of the first embodiment are shown by the same reference numbers, and detailed description thereof is omitted. FIG. 7(a) is a plan view of the overall structure for explaining the preventive maintenance/repair device 100, and FIG. 7(b) is a front view of the overall structure for explaining the preventive maintenance/repair device 100. FIG. 8(a) is a structural view for explaining the thruster driving part, and FIG. 8(b) is a structural view for explaining a maintenance/repair mechanism 108. As shown in FIGS. 7(a) and 7(b), there are disposed: a device body 123; a pair of holding mechanisms 104a and 104b connected to the device body 123, the holding mechanisms 104 and 104b being capable of holding the device body 123 on an outer circumferential surface of a cylindrical structure 19; and a traveling and driving part 102 disposed on the device body 123, the traveling and driving part 102 being capable of circumferentially moving along the outer circumferential surface of the cylindrical structure 19. Disposed on a lower part of the device body 123 is a maintenance/repair mechanism 108 that maintains and repairs the cylindrical structure 19. In addition, disposed on opposed side surfaces of the device body 123 are a pair of thruster driving parts 106a and 106b that moves the device body 123 in water. The holding mechanisms 104a and 104b respectively include: a pair of arms 103a and 103b each having a shape along the outer circumferential surface of the cylindrical structure 19; and guide rollers 105a and 105b disposed on distal ends of the pair of arms 103a and 103b. Respectively connected to the pair of arms 103a and 103b are arm cylinders 109a and 109b for driving the arms 103a and the 103b pneumatically or hydraulically. The arm 103a and the arm cylinder 109a, and the arm 103b and the arm cylinder 109b, are respectively connected by links 110a and 110b that respectively transmit driving forces of the arm cylinders 109a and 109b to the arms 103a and 103b. As shown in FIGS. 7(a) and 7(b), the traveling and driving part 102 has a traveling wheel 101 which is driven in rotation, while being in contact with the outer circumferential surface of the cylindrical structure 19. Other structures are substantially the same as those of the first embodiment shown in FIGS. 3(a) and 3(b). The traveling and driving part 102 is removably mounted on the device body 123. As shown in FIG. 8(a), the thruster driving parts 106a and 106b respectively include thrusters 114a and 114b, and motors 107a and 107b for driving the thrusters 114a and 114b. Pulleys 113a and 113b are connected to output shafts of the motors 107a and 107b, and pulleys 113j and 113k are connected to rotational shafts of the thrusters 114a and 114b. Intermediate pulleys 113c and 113d are disposed on positions near the pulleys 113a and 113b. Intermediate pulleys 113e and 113f are disposed on positions nearer the thrusters 114a and 114b to the intermediate pulleys 113c and 113d. Further, intermediate pulleys 113g and 113h are disposed on positions nearer to the thrusters 114a and 114b to the intermediate pulleys 113e and 113f. As shown in FIG. 8(a), the pulleys 113a and 113b on the side of the motors 107a and 107b and the intermediate pulleys 113c and 113d are connected to each other by belts 112a and 112b. The intermediate pulleys 113e and 113f and the intermediate pulleys 113g and 113h are connected to each other by belts 112c and 112d. The intermediate pulleys 113g and 113h and the pulleys 113j and 113k on the side of the thrusters are connected to each other by belts 112e and 112f. As shown in FIG. 8(b), the maintenance/repair mechanism 108 includes a housing 115, an equipment 116 that maintains and repairs the cylindrical structure 19, and a motor 117 that drives the equipment 116 in a longitudinal direction of the cylindrical structure 19. A gear 118a is connected to an output shaft of the motor 117. Connected to the gear 118a is a ball screw 119 via a gear 118b and a gear 118c. A nut 120 is engaged with the ball screw 119, and the nut 120 is connected to one end of a container case 122 for containing the equipment 116. Connected to the other end of the container case 122 is a slide guide 121 that makes slidable the container case 122 with respect to the housing 115. The maintenance/repair mechanism 108 is removably mounted on the device body 123. As shown in FIG. 8(b), the equipment 116 of the maintenance/repair mechanism 108 is formed of an ultrasonic flaw-detecting probe, such as a phased-array UT probe, which can ultrasonically detect a flaw such as a crack of the cylindrical structure 19 without contacting the cylindrical structure 19. In order to neutrally float the preventive maintenance/repair device 100 so as to move the preventive maintenance/repair device 100 in water by thrust force of the thruster driving parts 106a and 106b, a buoyant member (not shown) is located in one of the device body 123 and the maintenance/repair mechanism 108. As shown in FIGS. 7(a) and 7(b), arranged on an upper surface of the device body 123 are pendant fittings 111a 111b to which a rope and the like is fitted when the preventive maintenance/repair device 100 is lowered from the upper part of the reactor pressure vessel 1 in a hanging manner. According to this embodiment, as shown in FIG. 8(a), when the preventive maintenance/repair device 100 is moved in the water, the motors 107a and 107b of the thruster driving parts 106a and 106b are driven. The rotations of the motors 107a and 107b are transmitted to the intermediate pulleys 113c and 113d through the pulleys 113a and 113b and the belts 112a and 112b. Then, the rotations are transmitted to the intermediate pulleys 113g and 113h through the intermediate pulleys 113e and 113f connected to the intermediate pulleys 113c and 113d and through the belts 112c and 112d. The rotations are further transmitted to the pulleys 113j and 113k though the belts 112e and 112f connected to the intermediate pulleys 113g and 113h. Thus, the rotations of the motors 107a and the 107b are transmitted to the thrusters 114a and 114b, so that the thrusters 114a and the 114b are rotated. Since the buoyant member (not shown) is located in one of the device body 123 and the maintenance/repair mechanism 108, the preventive maintenance/repair device 100 can neutrally float. Thus, the preventive maintenance/repair device 100 can be moved in the water by the thrust force of the thruster driving parts 106a and 106b. As a result, the preventive maintenance/repair device 100 can be moved in the water, without the use of an access device that attaches/detaches the preventive maintenance/repair device 100 to/from the cylindrical structure 19. As shown in FIGS. 7(a) and 7(b), the traveling wheel 101 of the traveling and driving part 102 is in contact with the outer circumferential surface of the cylindrical structure 19. In addition, by driving the arm cylinders 109a and 109b of the holding mechanisms 104a and 104b, the guide rollers 105a and 105b disposed on the distal ends of the arms 103a and 103b are pressed onto the outer circumferential surface of the cylindrical structure 19 via the links 110a and 110b and the arms 103a and 103b. Therefore, the preventive maintenance/repair device 100 can be held on the outer circumferential surface of the cylindrical structure 19. As shown in FIG. 8(b), when the equipment 116 of the maintenance/repair mechanism 108 is moved in the longitudinal direction of the cylindrical structure 19, the motor 117 of the maintenance/repair mechanism 108 is driven. The equipment 116 is slidably disposed in the longitudinal direction of the cylindrical structure 19 with respect to the fixed housing 115 by the slide guide 121. The rotational drive of the motor 117 is transmitted to the ball screw 119 via the gears 118a, 118b, and 118c, and the rotational motion of the ball screw 119 is converted into a vertically linear motion by the nut 120. Thus, the equipment 116 connected to the nut 120 is slid in the longitudinal direction of the cylindrical structure 19. Accordingly, the equipment 116 can be precisely moved toward a target region of the cylindrical structure 19 in the longitudinal direction of the cylindrical structure 19. The traveling and driving mechanism 108 can be easily mounted on and removed from the device body 123. Thus, if the maintenance/repair mechanism 108 is broken for some reason or another, the whole preventive maintenance/repair device 100 is drawn upward, and the broken maintenance/repair mechanism 108 can be replaced with another normal maintenance/repair mechanism 108, which has been prepared beforehand, for a short period of time. As shown in FIGS. 7(a) and 7(b), when the preventive maintenance/repair device 100 is circumferentially moved along the outer circumferential surface of the cylindrical structure 19, the motor (not shown) of the traveling and driving part 102 is driven so that the traveling wheel 101 is driven in rotation. When the preventive maintenance/repair device 100 is circumferentially moved along the outer circumferential surface of the cylindrical structure 19, the guide rollers 105a and 105b disposed on the distal ends of the arms 103a and 103b of the holding mechanism 104a and 104b are rotated in accordance with the movement of the preventive maintenance/repair device 100. Also at this time, as described above, the guide rollers 105a and the 105b are pressed onto the outer circumferential surface of the cylindrical structure 19 by the arm cylinders 109a and 109b of the holding mechanisms 104a and 104b. Thus, the preventive maintenance/repair device 100 can be smoothly, circumferentially moved along the outer circumferential surface of the cylindrical structure 19, while the preventive maintenance/repair device 100 is being held on the outer circumferential surface of the cylindrical structure 19. Accordingly, the equipment 116 of the maintenance/repair mechanism 108 can be smoothly moved to a desired circumferential position along the outer circumferential surface of the cylindrical structure 19. As a result, the cylindrical structure 19 can be maintained and repaired over all the circumferential surface thereof by the equipment 116. The traveling and driving part 102 can be easily mounted on and removed from the device body 123. Thus, if the traveling and driving part 102 is broken for some reason or another, the whole preventive maintenance/repair device 100 is drawn upward, and the broken traveling and driving part 102 can be replaced with another normal traveling and driving part 102, which has been prepared beforehand, for a short period of time. When the equipment 116 of the maintenance/repair mechanism 108 shown in FIG. 8(b) is formed of an ultrasonic flaw-detecting probe, a target region of the cylindrical structure 19 can be ultrasonically detected. Thus, whether there is a crack or not in a welding line of the cylindrical structure 19 can be checked. In addition, various maintaining and repairing operations are possible by using other equipments 116 that are similar to the equipments in the first embodiment. Due to the provision of the pendent fittings 111a and 111b on the upper surface of the device body 123, the preventive maintenance/repair device 100 can be lowered from the upper part of the reactor pressure vessel 1 in a hanging manner, by connecting a rope or the like to the pendent fittings 111a and 111b. According to this embodiment, the preventive maintenance/repair device 100 can be securely held on the outer circumferential surface of the cylindrical structure 19 installed in the reactor pressure vessel 1. In addition, the preventive maintenance/repair device 100 can be circumferentially moved on the outer circumferential surface of the cylindrical structure 19. In addition, the preventive maintenance/repair device 100 can be moved in water. Thus, the preventive maintenance/repair device 100 can be precisely moved to a target region of the outer circumferential surface of the cylindrical structure 19, so that the outer circumferential surface of the cylindrical structure 19 can be maintained and repaired.
050135228	abstract	For determining the content and the chemical composition of particulate compounds in a liquid flow (15), the particulate compounds are collected on a filter (20) in a container (19) of a material which is permeable to microwaves. When carrying out the determination, samples of a fixed size of the liquid in the flow are supplied batchwise to the container on one side of the filter via an openable and a closable connection (16, 17) between the liquid flow and the container. At the same time, liquid from each liquid sample supplied batchwise, which has passed through the filter, is discharged from the container via an openable and a closable outlet (30, 31) on the other side of the filter. After closing of the connection between the liquid flow and the container and of the outlet for the liquid having passed through the filter, the container with the particulate compounds collected on the filter is supplied with a fixed amount of the solvent for the particulate compounds and the particulate compounds are dissolved in the solvent while being heated in a microwave oven. The solution of the particulate compounds, thus obtained, in each liquid sample supplied batchwise is led via an openable and a closable connection (33, 34) to an analysis apparatus (35) in which the composition and the content of one or more of the particulate compounds are determined. All actions, such as opening and closing of valves, heating processes and times for different operations, are controlled by automatic control.
044180365	abstract	A fuel assembly for a nuclear reactor comprises a 5.times.5 array of guide tubes in a generally 20.times.20 array of fuel elements. The guide tubes are arranged to accommodate either control rods or water displacer rods. The fuel assembly also comprises a plurality of Inconel and Zircaloy grids arranged to provide stability of the fuel elements and guide tubes while allowing the flow of reactor coolant therebetween.
	description	The present application is a division of U.S. patent application Ser. No. 11/165,972 which was filed on Jun. 24, 2005 and issued on Oct. 14, 2008 bearing U.S. Pat. No. 7,436,932, is assigned to the assignee of the present invention, and is incorporated by reference herein. X-ray radiation sources and, more particularly, X-ray radiation sources with low neutron emissions for radiation scanning of objects. X-ray radiation sources are commonly used in radiation inspection systems for non-destructive inspection of objects. X-ray radiation may be generated in such sources by the impact of a beam of accelerated electrons on a high atomic number (“Z”) target material, such as tungsten or tantalum. The electrons are accelerated by a potential difference established across a chamber, referred to as an acceleration energy. The deceleration of the incident electrons by the nuclei of the atoms of the target material generates radiation, referred to as Bremsstrahlung. A collimator is provided to direct some of the generated radiation onto an object to be inspected and to form the generated radiation into a beam of a desired size and shape. One or more radiation detectors are provided to measure radiation transmitted through and/or scattered from the object. The body of the radiation source may also be shielded. In order to prevent radiation from escaping the radiation inspection system, shielding is also provided around the system as a whole. Small objects, such as luggage and carry-on bags, are typically examined by radiation in the kilovolt range. However, radiation in the kilovolt range may not penetrate objects thicker than about 5 feet (1.52 meters), particularly if the object is filled with dense material. Standard cargo containers are typically 20-50 feet long (6.1-15.2 meters), 8 feet high (2.4 meters) and 6-9 feet wide (1.8-2.7 meters). Air cargo containers, which are used to contain a plurality of pieces of luggage or other cargo to be stored in the body of an airplane, may range in size (length, height, width (thickness)) from about 35×21×21 inches (0.89×0.53×0.53 meters) up to about 240×118×96 inches (6.1×3.0×2.4 meters). Large collections of objects, such as many pieces of luggage, may also be supported on a pallet. Pallets, which may have supporting sidewalls, may be of comparable sizes as cargo containers, at least when supporting objects. The term “cargo conveyance” is used to refer to all types of cargo containers and comparably sized pallets (and other such platforms) supporting objects. Higher energy radiation beams are required to penetrate through denser materials than less dense materials and through thicker materials than less thick materials. The low energies used in typical X-ray luggage and bag scanners, described above, are generally too low to penetrate through the much larger cargo containers, particularly those with widths or thicknesses of 5 feet (1.5 meters) or more. While the required energy level depends on the contents of the container and the width of the container, radiation in the megavolt range is typically required. 6 MeV to 10 MeV may be used, for example. 9 MeV is commonly used because it will penetrate through most cargo containers, regardless of the contents. However, high Z and medium Z metals commonly used in X-ray radiation sources, such as tungsten, tantalum, and molybdenum comprise stable isotopes having neutron production thresholds (the energy required to remove a neutron from a nucleus of an atom of the isotope) in a range of about 6 MeV to about 10 MeV. For example, the calculated neutron production thresholds for the stable isotopes of tungsten range from 6.191 MeV to 8.415 MeV. The calculated neutron production threshold for the stable isotope of tantalum is 7.651 MeV. The calculated neutron production thresholds for the stable isotopes of molybdenum range from 7.369 MeV to 12.667 MeV. Since 6 MeV to 10 MeV is a common range to examine cargo conveyances, neutrons are typically produced. Because of their ability to absorb larger amounts of photons than lower atomic number metals per unit volume, high Z metals, such as tungsten and lead, are also typically used to shield the target and collimate the radiation beam. The stable isotopes of lead have calculated neutron production thresholds of from 6.737 MeV to 8.394 MeV. If the generated X-ray radiation used to examine the objects has an energy above the neutron production threshold of the shielding material and collimator, neutrons will also be produced. Since neutrons may be harmful to people proximate the scanning system, thicker shielding may be required to prevent the escape of neutrons from the scanning system or the room containing the scanning system. This may increase the size and cost of the system. Concrete walls are commonly used to shield a room containing a scanning system, preventing or decreasing the amount of neutrons and X-rays that may escape the room. If space or other requirements prevent the use of concrete walls, then a multi-layer wall may be used. For example, a thick wall of polyethylene or borated polyethylene may be used as an inner layer to shield neutrons and lead or steel may be used as an outer layer to shield X-rays. The outer layer also shields gamma rays emitted by the polyethylene. Varian Medical Systems, Inc., (“Varian”) Palo Alto, Calif., sells X-ray radiation sources for medical therapy supported by a rotatable gantry that also supports a detector array. The gantry, the source, and the detector comprise an integrated unit, which is sold under the trade name CLINAC®. The radiation source comprises a copper target and tungsten shielding. Copper generates sufficient X-ray radiation for therapeutic purposes, and is less expensive than tungsten. CLINACs® are available at 4 MeV, 6 MeV, 10 MeV and above. Copper has two stable isotopes, copper-65, with a calculated neutron production threshold of 9.910 MeV, and copper-63, with a calculated neutron production threshold of 10.852 MeV. Since in the 10 MeV and above models of the CLINAC® the acceleration energy is above the neutron production threshold of copper-65 and tungsten, neutrons are produced. The collimator comprises a combination of tungsten and lead, which will also generate neutrons. The tungsten shielding will generate neutrons, as well. Efforts have been made to reduce neutron emission from X-ray sources used in medical therapy, such as radiation treatment. See, for example, Neutron Contamination from Medical Electron Accelerators, NRCP Report No. 79, National Council on Radiation Protection and Measurements, Bethesda, Md., pp. 59-60 (1995). It is said to be difficult to reduce the number of neutrons produced per useful photon rad of radiation, in the space available in existing sources. (Id.) It is noted that neutron emission may be reduced by absorbing unwanted neutrons in a medium Z material, such as iron, instead of tungsten or lead; but it is also noted that much more iron is required than tungsten or lead and there is insufficient space to take advantage of this reduction completely. (Id.) Varian also sells a Linatron® series of X-ray sources that generate X-ray radiation in the range of 1-10 MeV. In these sources, the target is tungsten, typically in the form of a disk. A disk of copper is attached to the downstream side of the tungsten, of the electron beam, to dissipate heat and to act as a final electron stop for electrons passing through the tungsten target. The tungsten target is the primary source of X-ray radiation. It is believed that a small amount of X-ray radiation may be generated by the copper disk as well, by the electrons passing through the tungsten target. If the acceleration energy of the source is greater than the neutron production threshold of the tungsten, neutrons may be produced. In accordance with an embodiment of the invention, a method of manufacturing a radiation source is disclosed comprising selecting for at least one of a target, a collimator, and target shielding (in other words, the target, the collimator, and/or the target shielding) at least one isotope having a neutron production threshold greater than a peak acceleration energy of the source and assembling the source including the selected material. The method may further comprise selecting housing shielding material consisting essentially of at least one isotope having a neutron production threshold greater than the peak acceleration energy and assembling the source with the selected housing shielding material. The method may further comprise preparing a preliminary design for a radiation source to meet, at least, neutron production requirements, inputting the preliminary design into a simulation to predict neutron production, and receiving an output of the simulation. If the output does not meet requirements, the method may further comprise adjusting the design and inputting the adjusted design into the simulation to predict neutron production. The method may further comprise preparing the preliminary design based on size requirements and/or X-ray radiation generation requirements. As used herein, the term “about” refers to differences due to round off errors and typical measurement capabilities; the term “at least one of” means “any one or more of the following”; the term “consisting essentially of” means that isotopes of the same and/or other materials that will not generate neutrons at the peak acceleration energy of the source, may be included; and the term “peak acceleration energy” means “maximum” acceleration energy. In accordance with embodiments of the invention, radiation sources, such as X-ray radiation sources, provide no neutron production or reduced neutron production as compared to radiation sources comprising typical materials for the source, collimator, and/or shielding by using materials with neutron production thresholds above the peak acceleration energy of the source. (As mentioned above, here, the term “peak acceleration energy” means “maximum” acceleration energy.) For example, an X-ray source with a peak acceleration energy of less than about 9.9 MeV, comprising a copper target, a copper collimator, and copper shielding of the target and housing, will generate no neutrons. Where it is not feasible to use only materials with neutron production thresholds below the peak acceleration energy of the source due to size, weight, and/or cost constraints, neutron production may be reduced by using such materials for all or part of certain components. By proper selection of materials, no or reduced neutron production may be provided in radiation sources with peak acceleration energies, and hence peak radiation energies, in the range of from about 6.2 MeV to less than about 13.1 MeV. One of the advantages of reduced or no neutron production is that the physical size of the shielding of the source and the scanning system as a whole, including the room containing the scanning system, may be reduced. Also, the risk of activation of certain materials of or within the object under test and the scanning room by neutron capture reactions, may also be reduced. The class of materials that may be used is referred to herein as “low atomic number materials” or “low Z materials” because those materials have atomic numbers (“Z”) significantly less than that of tungsten (Z=74), tantalum (Z=73), lead (Z=82), and molybdenum (42), the common materials used in the prior art. Copper, for example, has an atomic number of 29. In one example, the low Z material may have an atomic number Z of 30 or less. Appropriate low atomic number materials for a particular application have their lowest neutron production threshold greater than the peak acceleration energy of the source, so that no neutrons are generated by that material. It has been found that due to their size differences, more neutrons may be generated by the shielding and collimator materials than by the target. It is also noted that not all low Z materials are appropriate in the preferred peak acceleration range for examining cargo conveyances of from about 6 MeV to 10 MeV, which is a useful range for examining cargo conveyances, which can be greater than 5 feet (1.524 meters) thick. Beryllium, for example, with an atomic number of 4, has the lowest neutron production threshold of all elements at 1.665 MeV. FIG. 1 is a schematic cross-sectional view of an example of a radiation source 100, such as an X-ray linear accelerator, in accordance with an embodiment of the present invention. The linear accelerator 100 comprises a housing 110 defining an acceleration chamber 120. The housing defines an input 125 and an output 130. A target 140 is supported in or near the output 130 of the housing 110. The target 140 is a material that generates Bremsstrahlung radiation in response to impact by accelerated charged particles. In one example, the target 140 preferably consists essentially of isotopes of a low Z material having a neutron production threshold greater than the acceleration potential of the acceleration chamber 120. Copper may be used, for example. An electron gun 150 extends through the input 125. The electron gun comprises a filament 160, which is suspended in the acceleration chamber 120. A magnetron 165 is coupled to the acceleration chamber 120 to create an electromagnetic field within the acceleration chamber. The electromagnetic field accelerates electrons generated by the filament 160 to a desired energy level within the acceleration chamber 120. The accelerated electrons form an electron beam 170 that strikes the target 140 in the output 130, causing the emission of photons in the form of a beam of X-ray radiation 175. The housing 110 may comprise a thin copper wall, for example. In this example, the target shielding material 180 circumferentially surrounds the target 140, to prevent the escape of X-ray radiation in a direction perpendicular to the electron beam. Housing shielding 182 may be provided around the housing 110, if needed. The need for housing shielding 182 may depend on the accelerator design, the target shielding 180, and the required level of attenuation. If the accelerator uses a solenoid or otherwise narrowly focuses the electron beam on the target 140, and/or required attenuation is low, housing shielding may not be needed to shield stray electrons. If the target shielding 180 extends sufficiently behind the target, then the housing shielding 182 may not be needed to shield radiation emitted behind the target. The embodiment of FIGS. 4-6 shows such a design. In accordance with this embodiment of the invention, all or a portion of the target shielding 180 and the housing shielding 182, if present, also comprise isotopes of appropriate low Z materials having neutron production thresholds greater than the peak acceleration energy of the source, to avoid the generation of neutrons by the shielding, although that is not required. Copper may be used, for example. The target shielding 180 and the housing shielding 182 may be a single piece or one or more separate pieces of material. The shielding materials 180 and 182 may be the same or different. A collimator 190 is coupled to a distal end of the housing 110. It may be connected to the target shielding 182 and to the housing shielding 182, for example. The material of the collimator 190 defines a passage 195 to allow passage of the radiation 175. The passage is shaped to define the radiation beam 175. In accordance with this embodiment, all or a portion of the collimator 190 is composed of isotopes of a low Z material to avoid the generation of neutrons, although that is also not required. Copper may be used, for example. The collimator 190 may also comprise a single piece of material or multiple pieces. The thicknesses of the shielding material 180, 182 and the collimator 190 may vary in different locations on the linear accelerator 100. At peak energies above 1 MeV, the intensity of the radiation emitted from the target is greatest in the forward direction, along the axis of the path of the electron beam 170 through the target 140. The intensity decreases as the angle from the axis increases. The collimator 190 may therefore be thicker than the shielding material 180, 182. The shielding material 180, 182 may also be thicker at angles closer to the axis, as is known in the art. When the target 140 is made of an appropriate low Z material, with a neutron production threshold above the peak acceleration energy of the source, the shielding 180, 182 and the collimator 190 need not be as thick as when the target is a high Z material because neutrons need not be shielded. The different components of the source 100 may comprise different materials. For example, a natural copper target 140, a natural copper shielding 180a, 180b, and a natural iron collimator 190 may be used in an X-ray source 100 operating at a peak energy less than the neutron production threshold of iron-57 of 7.646 MeV. A low Z material may have different stable isotopes with different neutron production thresholds. The isotopic composition of the low Z material therefore has to be considered when selecting appropriate materials for the target 140, shielding 180, 182, and collimator 190 to achieve a desired reduction or elimination of neutrons. To avoid all neutron production, the target 140, shielding 180, 182 and collimating materials 190 are selected such that the neutron production threshold for all isotopes exceeds the peak energy of the particular radiation source. In one example, naturally occurring copper comprises 69.17% of copper-63, which has a calculated neutron production threshold of 10.852 MeV, and 30.83% of copper-65, which has a calculated neutron production threshold of 9.910 MeV. Naturally occurring copper may therefore be used as a target 140, shielding 180, 182, and collimator 190 material for sources operating at a peak acceleration energy below 9.910 MeV, for zero neutron production. Isotopically pure copper-63 can be used as a target 140, shielding 180, 182, and collimator 190 material for sources operating at a peak acceleration energy below 10.852 MeV. In another example, naturally occurring iron comprises 91.75% iron-56, having a calculated neutron production threshold of 11.197 MeV, 5.85% iron-54, having a calculated neutron production threshold of 13.37 MeV, 2.12% iron-57, having a calculated neutron production threshold of 7.646 MeV, and 0.28% iron-58, having a calculated neutron production threshold of 10.044 MeV. Hence, if naturally-occurring iron is used as a target 140, shielding 180, 182, and collimator 190, an X-ray source operating at a peak acceleration energy of less than 7.646 MeV will not generate neutrons. Isotopically pure iron shielding consisting of only iron-56 could be used as a target 140, shielding 180a, 180b, and collimator 190 in X-ray sources 100 operating at peak acceleration energies up to 11.197 MeV without generating neutrons. It is noted that if an iron target is used in an X-ray source operating up to (but less than) 10.044 MeV, only about 2.12% of the target 140, shielding 180, 182, and collimator 190 would generate neutrons. Therefore, at a peak energy of less than 10.044 MeV, a naturally occurring iron target 140, shielding 180, 182, and collimator 190 would provide a significantly reduced amount of neutron production, as compared to the use of tungsten, but would not eliminate it. If some amount of neutron generation may be tolerated, but less neutron generation is desired than if tungsten, tantalum, molybdenum, or lead are used, metals may be chosen that provide certain isotopes having neutron production thresholds below the peak energy of the source and certain isotopes having neutron production thresholds above the peak energy of the source. For example, in a source having a peak acceleration energy of 8.5 MeV, the following materials may be used for the target 140, the shielding 180, 182, and the collimator 190: magnesium (10% of which consists of magnesium-24 having a calculated neutron production threshold of 7.331 MeV), iron (2.12% of which consists of iron-57 having a calculated neutron production threshold of 7.646 MeV), nickel (3.63% of which consists of nickel-61 having a calculated neutron production threshold of 7.82 MeV), and zinc (4.10% of which consists of zinc-67 having a calculated neutron production threshold of 7.051 MeV). Different materials from this group may be used for the target 140, the shielding 180, 182, and the collimator 190. Such reductions could also be advantageous in particular applications. Shielding requirements may be reduced, for example. In another example of a configuration that will reduce but not necessarily completely eliminate neutron production, certain components, such as the target 140, for example, may be an appropriate low Z material, while one or more other components, such as the collimator 190, the target shielding 180, and/or the housing shielding 182 may comprise high Z materials that may generate neutrons, such as tungsten, tantalum or lead, for example. Use of tungsten, tantalum, lead, or other such materials may provide better X-ray radiation generation or shielding, and may therefore be needed for certain components, or part of certain components, to meet performance and size requirements for the source, for example, as discussed below. Therefore, in accordance with another embodiment of the invention, neutron production is reduced by use of an appropriate low Z material for at least one but not necessarily all components, as compared to the use of tungsten, tantalum, molybdenum, or lead for that component. Use of isotopically purified high Z atom material as a target 140, shielding 180, 182, and/or collimator 190 may also decrease or eliminate neutron generation, and provide some or all of the benefits of the use of a high Z material. For example, isotopically pure tungsten-182, which has a calculated neutron production threshold of 8.064 MeV, can be used in a source having a peak acceleration energy less than 8.064 MeV, without producing neutrons. Isotopically pure molybdenum-94, which has a peak acceleration energy of 9.677 MeV, can be used in a source having a peak acceleration energy less than 9.677 MeV, without producing neutrons. Mixtures of appropriate isotopes that are above the peak acceleration energy can be used, as well. For example, a mixture of molybdenum-96 and molybdenum-94 can be used in a source having a peak acceleration energy less than 8.064 MeV without producing neutrons. While typically a metal, the target 140, shielding 180, 182, and/or the collimator 190 may be a non-metal, such as carbon. One stable isotope of carbon, carbon-13, which has an abundance of only 1.11%, has a calculated neutron production threshold of 8.071 MeV. The other stable isotope of carbon, carbon-12, which has an abundance of 98.89%, has a calculated neutron production threshold of 18.721 MeV. A source with a peak acceleration energy of less than 8.071 MeV would generate no neutrons, while in a source operating at above 8.071 and less than 18.721, only 1.11% of the carbon (the carbon-13) would generate neutrons. Preferably, a stable form of carbon is used, such as graphite or diamond, for example. As is known in the art, the neutron production threshold for every isotope of a metal, including low atomic number metals, can be calculated according to the following equation:Threshold(MeV)=(Mass Excess(Z,A-1)+Neutron Mass Excess)−Mass Excess(Z,A). In the equation above, Z is an atomic number of an atom of an element; A is a mass number (sum of the number of protons and neutrons) of an atom of the element; (Z, A) is an original isotope of an atom of the element (before it loses a neutron as a result of interacting with an X-ray radiation); and (Z, A-1) is a mass of an atom of a resulting isotope of the element after loss of one neutron. Mass Excess (Z, A-1) is the equivalent energy of the mass of the resulting isotope, as compared to carbon-12, which has a mass excess of 0 MeV. The Neutron Mass Excess is a constant equal to 8.071 MeV, which is the energy of the difference between the neutron mass of a particular isotope and the neutron mass of carbon-12, which is set at 0 MeV. The Mass Excess (Z, A) is the equivalent energy of the mass excess of the initial isotope. Mass instead of Mass Excess may be used to determine the threshold as well, as is known in the art. For example, the neutron production threshold for magnesium-24 is calculated as follows. Magnesium-24 is an original isotope which, after interaction with a beam of X-ray photons, emits one neutron and becomes magnesium-23, a resulting isotope. The mass excess (Z, A-1) of magnesium-23 (−5.473 MeV) is added to neutron mass excess (8.071 MeV). The mass excess (Z, A) of magnesium-24 is then subtracted (−13.933 MeV), yielding a neutron production threshold of 16.531 MeV, as shown below:(−5.473 MeV+8.071 MeV)−(−13.933 MeV)=16.531 MeV. Large negative mass excesses (Z, A) show that the protons and neutrons in a nucleus of an atom are tightly bound together. An amount of external energy exceeding the absolute value of the negative mass excess is necessary to remove one neutron from the nucleus. In other words, the neutron production threshold of a particular isotope is a minimum energy a photon needs to have to emit one neutron from a nucleus of that isotope. Calculated neutron production thresholds for each isotope of several materials are summarized in Table I, below. In Table I, some abundances may not add to 100%, due to round off errors. Abundance information is not provided for unstable isotopes. Neutron production thresholds for isotopes of elements ranging from beryllium to uranium may also be found in Neutron Contamination from Medical Electron Accelerators, National Council on Radiation Protection and Measurements, Bethesda, Md., pp. 18-23 (1995), where they are referred to as “separation energies.” TABLE INeutronMassProductionAtomicExcessThresholdAbundanceElement/IsotopeNumber(MeV)(MeV)(%)Beryllium-844.942Beryllium-9411.3481.665100.00%Carbon-11610.650Carbon-126018.72198.89%Carbon-1363.1258.0711.11%100.00%Magnesium-2312−5.473Magnesium-2412−13.93316.53178.99%Magnesium-2512−13.1937.33110.00%Magnesium-2612−16.21511.09311.01%100.00%Aluminum-2613−12.210Aluminum-27−17.19713.058100.00%Scandium-4421−37.816Scandium-4521−41.06911.324100.00%Titanium-4522−39.007Titanium-4622−44.12513.1898.25%Titanium-4722−44.9328.8787.44%Titanium-4822−48.48711.62673.72%Titanium-4922−48.5588.1425.41%Titanium-5022−51.42610.9395.18%100.00%Vanadium-4923−47.956Vanadium-5023−49.2189.3330.25%Vanadium-5123−52.19811.05199.75%100.00%Chromium-4924−45.326Chromium-5024−50.255134.35%Chromium-5124−51.445Chromium-5224−55.41312.03983.79%Chromium-5324−55.2817.9399.50%Chromium-5424−56.9299.7192.37%100.01%Manganese-5425−55.55225−57.70710.226100.00%Iron-5326−50.941Iron-5426−56.24913.3795.85%Iron-5526−57.475Iron-5626−60.60111.19791.75%Iron-5726−60.1767.6462.12%Iron-5826−62.14910.0440.28%99.99%Cobalt-5827−59.842Cobalt-5927−62.22410.453100.00%Nickel-5728−56.076Nickel-5828−60.22312.21868.08%Nickel-5928−61.1518.99926.22%Nickel-6028−64.46811.3881.14%Nickel-6128−64.2177.823.63%Nickel-6228−66.74310.5970.93%100.00%Copper-6229−62.795Copper-6329−65.57610.85269.17%Copper-6429−65.421Copper-6529−67.2609.91030.83%100.00%Zinc--6330−62.210Zinc--6430−66.00011.86148.60%Zinc--6530−65.908Zinc--6630−68.89711.0627.90%Zinc--6730−67.8777.0514.10%Zinc--6830−70.00410.19818.80%Zinc--6930−68.415Zinc--70−69.5609.2160.60%100.00%Molybdenum-9142−82.210Molybdenum-9242−86.80612.66714.84%Molybdenum-9342−86.805Molybdenum-9442−88.4119.6779.25%Molybdenum-9542−87.7097.36915.92%Molybdenum-9642−88.7929.15416.68%Molybdenum-9742−87.5426.8219.55%Molybdenum-9842−88.1138.64224.13%Molybdenum-9942−85.967Molybdenum-10042−86.1858.2899.63%100.00%Tantalum-18073−48.861Tantalum-18173−48.4417.65199.99%Tungsten -17974−49.300Tungsten -18074−49.6448.4150.12%Tungsten -18174−48.253Tungsten -18274−48.2468.06426.50%Tungsten -18374−46.3666.19114.31%Tungsten -18474−45.7067.41130.64%Tungsten -18574−43.389Tungsten -18674−42.5127.19428.43%100.00%Lead - 20382−24.805Lead - 20482−25.1248.3941.40%Lead - 20582−23.784Lead - 20682−23.8018.08824.10%Lead - 20782−22.4676.73722.10%Lead - 20882−21.7647.36852.40%100.00% FIG. 2 is a graph showing the lowest neutron production thresholds among isotopes of a number of materials presented in the table above. In FIG. 2, the neutron production thresholds from Table I are rounded off to the nearest tenth. In an X-ray source 100 comprising a target 140, shielding 180, 182, and a collimator 190 of a selected material (naturally occurring), whose peak acceleration energy is below each of these thresholds, no neutrons would be produced. For example, the lowest calculated neutron production threshold for copper is 9.910 MeV. Any source with a copper target 140, shielding 180, 182, and collimator 190 operating below 9.910 MeV would not generate neutrons during operation. In another example, as shown in the table above and in FIG. 2, if no or reduced neutron production (as compared to tungsten) is desired in an X-ray source 100 operating at a peak acceleration energy up to 7 MeV, the following naturally occurring low Z materials are appropriate for the target 140, the shielding 180, 182, and/or the collimator 190: carbon, magnesium, aluminum, scandium, titanium, vanadium, chromium, manganese, iron, cobalt, nickel, copper, and zinc. In another example, if no or reduced neutron production (as compared to tungsten) is desired in an X-ray source 100 with a peak acceleration energy in a range of from 7 MeV to 8 MeV, the following naturally occurring low Z materials are appropriate for the target 140, the shielding 180, 182, and/or the collimator 190: carbon, aluminum, scandium, titanium, vanadium, manganese, cobalt, and copper. In another example, if no or reduced neutron production (as compared to tungsten) is desired in an X-ray source 100 with a peak acceleration energy in a range of from 8 MeV to 9 MeV, the following naturally occurring low Z materials may be used for the target 140, the shielding 180, 182, and/or the collimator 190: aluminum, scandium, vanadium, manganese, cobalt, and copper. Radiation sources for use in scanning cargo conveyances often have a peak acceleration energy of 9 MeV to generate radiation with a peak energy of 9 MeV. 9 MeV is sufficient to penetrate most cargo conveyances, including standard cargo conveyances, regardless of the contents. Copper is a preferred material for use at peak acceleration energies of less than 9.910 MeV, including 9 MeV. In another example, if no or reduced neutron production (as compared to tungsten) is desired in an X-ray source 100 with a peak acceleration energy in a range of from 9 MeV to 10 MeV, the following naturally occurring materials may be used for the target 140, the shielding 180, 182, and/or the collimator 190: aluminum, scandium, and cobalt. At a peak acceleration energy in a range of from 10 MeV to 11 MeV, scandium and aluminum may be used and at a peak acceleration energy in a range of from 11 MeV to 13 MeV, aluminum may be used. In addition to the neutron production threshold, manufacturability and performance are other considerations in the selection of a target, collimator, and/or shielding materials. For example, the thermal conductivity, melting point, fatigue, ability to braze and bond, ability to vacuum seal, of particular materials, are important considerations. For target materials, the ability to generate Bremsstrahlung X-ray radiation and the quantity generated, are also a consideration. Copper meets many of these considerations, and is, therefore, a preferred low Z material. For example, the target may be GlidCop® AL-60 Dispersion Strengthened Copper, available from OMG Americas, Newark, N.J., for example. GlidCop® AL-60 is said to comprise 98.9% by weight naturally occurring copper and 1.1% by weight aluminum oxide (Al2O3). It is said to have has a thermal conductivity of 322 watts/meter-kelvin, a tensile strength of 413-517 MPa, and electrical resistivity of 2.2 e-006 ohm-cm. Oxygen-free Electronic Copper, UNS C10100, which is available from Hitachi Metals, Ltd, Japan, and Copper and Brass Sales, Southfield, Mich., for example, may be used for target shielding 220. UNS C101000 is said to be at least 99.99% by weight copper. Its density is said to be 8.89-8.94 g/cc and its Vickers hardness is said to be 75-90. A lower grade copper may be used for the collimator 190. Since low Z materials typically absorb fewer X-ray photons than an equal volume of the high Z materials, a greater volume of low Z shielding is necessary to absorb the same amount of photons. However, since the low Z materials have lower density than high Z materials, the weight of the radiation inspection system may remain almost the same as a system with the high Z material shielding. The thickness of shielding necessary to absorb a particular amount of photons having a particular energy level is calculated according to the following equation, for a point source emitting isotropically:I(t)=I0exp(−μ*t)/(4πR2),where t is the thickness of a shielding material between the source and the measurement point, I(t) is the intensity of the radiation after passing through the shielding material having the thickness t, I0 is the initial intensity of the radiation, μ is the X-ray attenuation coefficient, and R is the distance between the source and measurement point. As is known in the art, a broadbeam tenth value layer table for the material comprising the shield may be used to calculate μ for the material, at a particular energy. When an appropriate low Z material is used as the target 140 of the radiation source 100 to prevent or reduce neutron production, the probability of X-ray photon emission will be less than that of a high Z material, for each accelerated electron that hits the target. To compensate for this natural phenomenon, more electrons may be accelerated into the target 140. The absorption of neutrons by low Z shielding may reduce or eliminate the number of neutrons available to activate materials (generate X-ray radiation) due to neutron capture by other components of the X-ray scanning system, depending on the shielding material, as is known in the art. Shielding a radiation inspection system with a neutron-absorbing material facilitates use of such systems as mobile systems as opposed to stationary systems of prior art. FIG. 3 is an example of a method 200 of manufacturing an X-ray radiation source with no neutron production in accordance with an embodiment of the invention. A desired peak acceleration energy and/or peak radiation energy of a source is selected, in Step 210. A target material having a neutron production threshold greater than the peak acceleration energy or the peak radiation energy is selected, in Step 220. Shielding material having a neutron production threshold greater than the acceleration energy is selected, in Step 230. A collimator material having a neutron production threshold greater than the acceleration energy is selected, in Step 240. The source is assembled with the selected material or materials, in Step 210. Steps 220, 230 and 240 may be selected in any order. The same or different materials may be selected for each component. The material for each component may be selected based on a table of neutron production thresholds such as the table of Table I, a graph as in FIG. 2, or calculations, as described above, for example. If there are no size, weight, and/or cost constraints with respect to the X-ray source 100, the target 140, shielding 180, 182 and the collimator 190 may all consist essentially of a material or materials whose neutron production threshold is greater than the peak acceleration energy of the source. However, the amount of low Z material, such as copper, needed to provide X-ray and neutron shielding and X-ray collimation, for example, may take up a volume about twice as large as tungsten, and would cost more than the use of tungsten. The size, weight, and cost differences for other low Z materials are comparable. In a particular application, size, weight, and cost considerations may therefore need to be balanced against the value of a reduction or elimination of neutron production. Materials both above and below the neutron production threshold may be needed to meet the requirements for X-ray production, neutron production, size, weight, and/or cost in a particular application. FIG. 4 shows an example of a cylindrical X-ray head 300 for use in an X-ray source in accordance with an embodiment of the invention, designed to meet predetermined neutron production and size requirements in a source with a predetermined acceleration energy. For example, the X-ray head may be designed to produce up to a predetermined amount of neutrons per hour (which is less than would be produced if tungsten were used) at a peak acceleration energy of 9 MeV, and to encompass a predetermined volume (which is less than the volume required if the target, shielding and collimator were to be made of only copper). To meet the volume and neutron production requirements, the X-ray head comprises a combination of copper, tungsten, and lead. In FIG. 4, a collimator 305 is shown surrounded by lead shielding 309. In this example, the collimator defines a passage 306 to define a fan beam of radiation. FIG. 5 is a cross-sectional, perspective view of the X-ray head 300 of FIG. 4. The X-ray head 300 comprises a copper target disk 315 within a target assembly 317. The target assembly 317 also comprises a drift tube 319 extending from a guide front end 321 to the target 315. The guide front end 321 is coupled to an output of an acceleration chamber of a linear accelerator, for example, in use. Electrons accelerated by an acceleration chamber enter the drift tube 319 and collide with the target 315. The target 315 may be thick enough to stop all electrons. A target assembly shielding 324 is provided around the target assembly 317 to attenuate the escape of X-ray radiation from the target 315 in directions perpendicular to the electron beam and behind the target. The target assembly shielding 324 comprises a first target assembly shielding section 326 of copper in the form of a cylinder around the target assembly 317, and a second target assembly shielding section 328 of tungsten behind the target assembly 317. In this example, the first target assembly shielding section 326 comprises a tenth value layer (“TVL”) of copper, which is sufficient to attenuate X-ray radiation by 10%. Additional sections of different materials may be provided, as well. The collimator 305 shown in FIG. 4 is downstream of the target 315 in FIG. 5. The collimator 305 comprises a first, upstream collimator section 330 of copper and a second, downstream collimator section 332 of tungsten. The upstream collimator section 330 also comprises a TVL of copper. The first and second sections 330, 332 are cylinders with matching outwardly tapered inner diameters forming the passage 307, also shown in FIG. 4, shaped to define the radiation beam in this example. The second collimator section 332 may comprise a plurality of tungsten disks for ease of manufacture and assembly. Multiple passages may be defined, as well. Additional sections of different material may be provided, as well. The passage 307 may be shaped to define a cone beam, a pencil beam, or other such desired shape, as well. The collimator 305 impedes the escape of X-ray radiation in directions forward of the target 315, outside of the passage 307. The lead shielding 310, also shown in FIG. 4, encases the target assembly shielding 217, the collimator 305, and the upstream target assembly shielding 324. Sufficient tungsten in the second target assembly shielding section 32 and in the second collimator section 332, as well as lead shielding 310 are provided to meet space and X-ray leakage requirements for a particular application. FIG. 6 is an enlarged cross-sectional view of the target assembly 317 of FIG. 5, showing the copper disk 315 and the drift tube 319. Cooling tubes 340 are also shown. Openings 342 are provided in the target 315 to receive cooling fluid provided by the cooling tubes 340, as is known in the art. A downstream end of the drift tube 319 extends partially into the target 315. A metallic vacuum sealing flange 344, such as a Conflat® mount, available from Varian Inc., Palo Alto, Calif., or MDL Vacuum Products Corporation, Hayward Calif. is provided at one end of the drift tube, to form a vacuum seal in the target assembly 317, as is known in the art. A target insulator 346 is provided around the drift within the first target assembly shielding section 326 of copper tube 319. A flexure 348 of soft material, such as copper is also provided around the drift tube 319, to compensate for tolerance errors, as is known in the art. The X-ray head 300 is coupled to the forward end of a linear accelerator, for example. An example of suitable linear accelerator 100 is shown in FIG. 1. The X-ray head 300 may be attached to the downstream end of the linear accelerator 100, instead of the target 140 and target shielding 180. An upstream end of the drift tube 319 is coupled to an output of the acceleration chamber, such as the acceleration chamber 120 in FIG. 1, with the desired acceleration energy. During operation, electrons generated by the electron gun 150 are accelerated, as discussed above. The accelerated electrons enter the drift tube 319 and impact the copper target 315. Bremsstrahlung X-ray radiation with a peak energy of 9 MeV is generated as the copper target 315 decelerates the electrons of the electron beam. Since the peak acceleration energy of the linear accelerator 100 (9 MeV) is less than the lowest neutron production threshold of the copper target 315 (9.910 MeV), no neutrons will be produced by the target 315. X-ray radiation emitted forward of the target 315 is collimated into a desired shape, here a fan beam, by the passage 307 (see FIGS. 4 and 5). Much of the forward emitted radiation that does not pass through the passage 307 is absorbed by the first collimator section 330 of copper. Since copper has a neutron production threshold greater than the peak acceleration energy of the source and of the generated X-ray radiation, no neutrons are produced. Much of the remaining radiation is absorbed by the second collimator section 332 of tungsten. While that X-ray radiation may cause the generation of neutrons by the second collimator section 340 of tungsten, since the amount of X-ray radiation reaching the second section has already been reduced by the first section 335, many fewer neutrons are produced than if the entire collimator were to be of tungsten, tantalum, or lead, for example, as in the prior art. If the entire collimator 305 is copper, while no neutrons would be generated, to have the same shielding efficiency, for X-ray radiation, the collimator 305 would have to be much larger. As indicated above, about 10% of the X-ray radiation emitted by the target 315 towards the first target assembly shielding 326 of copper will be absorbed by the TVL of copper shielding. Neutrons will not be generated, for the same reason discussed above. Much of the radiation that passes through the target assembly shielding 326 will be absorbed by the lead housing 310 and the second target assembly shielding 328 of tungsten. Small leakage may be tolerated. While neutrons may be generated by the lead and tungsten, which have neutron production thresholds below the peak energy of the radiation, the X-ray radiation is so attenuated, that the neutron generation will be low—much lower than if the target and shielding were all tungsten, tantalum, or lead. In addition, the radiation emitted to the rear of the target 315 is of much lower intensity than that emitted forward of the target. While copper could be used in the rear target assembly shielding instead of tungsten, much more copper would be required to achieve the same shielding efficiency as the tungsten, as discussed above. An overall reduction in neutron production of up to 100% may be achieved depending on the allowable space. It is noted that due to the design of the X-ray head 300, housing shielding, such as housing shielding 182 in FIG. 1, may not be required to shield X-ray radiation emitted in a direction behind the target 315, because sufficient shielding is provided by the X-ray head 300. The accelerator design may also obviate the need for housing shielding to shield stray electrons, as discussed above. If housing shielding is used, it may also be an appropriate low Z material, such as copper, for example. In addition, a combination of a low Z material, such as copper, and a high Z material, such as tungsten, may also be provided if needed in a particular application. Isotopes of tungsten and molybdenum that will not generate neutrons at the peak acceleration energy of the source may also be used instead of or along with copper and other materials in the target 315 and the other components of the X-ray head 300. A source 100 or X-ray head 300 may be developed to meet acceleration energy, X-ray production, neutron production, size, weight, and cost requirements with the assistance of a Monte Carlo or other such random event simulation, in an iterative process. An MCNP5 Monte Carlo simulation, available from Oak Ridge National Laboratories, Oak Ridge, Tenn., for example, may be used. Other source configurations to meet other requirements may be similarly developed by those skilled in the art with Monte Carlo or other such simulations, and the teachings of the present invention. An example of an iterative process 400 to design a source in accordance with another embodiment of the invention is shown in FIG. 7. A first, preliminary design is prepared to meet maximum neutron production requirements, in Step 410. The design may be made to meet other requirements as well, such as X-ray production levels, size, and weight, for example. Assuming that it is known that use of copper for the target, collimator and shielding will result in a source greater than the allowable size, and that predetermined amount of neutron production is acceptable, the first, preliminary design may comprise a copper target, a tungsten target assembly shielding, a tungsten collimator, a tungsten upstream target assembly shielding, and lead shielding, for example, of predetermined sizes. A simulation, such as a Monte Carlo simulation, is run, in Step 420, to determine the expected neutron production resulting from the preliminary design. The expected level of X-ray production may also be determined. The results of the simulation are evaluated, in Step 430. If the requirements are met, the source may be assembled based on the design, in Step 440. If the requirements are not met, then the design is revised based on the results of the simulation, in Step 450. For example, if the results of the simulation in Step 440 show that there would be too much neutron production, then all or part of the tungsten target assembly shielding may be made of copper. The simulation is run again, in Step 420, and the results evaluated, in Step 430. If the simulation shows that the neutron production is still too high, another revised design may be prepared. In the next revised design, a portion of the collimator closest to the target may be made of copper for example. Steps 420-450 may be repeated until the neutron production requirements and other requirements, if present, are met. The simulation facilitates testing of minor adjustments in design and dimensions to achieve a final design, as well. A radiation source in accordance with embodiments of the invention may be used in a radiation inspection system, such as a cargo scanning system 500 depicted in FIG. 8, for example, in accordance with another embodiment of the invention. An X-ray source 502, which may be similar to the source 100 of FIG. 1 and/or may include the X-ray head 300 of FIGS. 4-6, for example, is shown on one side of a cargo conveyance 504. The cargo conveyance 504 may be a standard cargo conveyance, which has a width of about 6-9 feet (1.8-2.7 meters) or other sized cargo conveyance. The source and system of embodiments of the invention may be particularly useful with cargo conveyances with thicknesses of 5 feet (1.5 m) or more, since higher energy radiation beams are required to penetrate the thickness. For example, the peak acceleration energy of the source may be at least about 6.2 MeV, at least about 6.7 MeV, or at least about 7.7 MeV, to generate radiation having a peak (maximum) energy of at least about 6.2 MeV, 6.7 MeV, or 7.7 MeV, respectively, depending on the configuration. The peak acceleration energy may be about 9 MeV, for example. The source 502 includes a collimator 503. A slot 503a is provided through the collimator 503 to define the radiation beam R. A detector 506 is supported on an opposite side of the cargo conveyance 504, to detect radiation interacting with the cargo conveyance. The detector 506 may be positioned to detect radiation transmitted through the cargo conveyance 504, for example. The cargo container 504 is conveyed by a conveyor system 508 through a shielded tunnel 510, between the source 502 and the detector 506. The detector 506 may be an L-shaped detector array, with a first arm 512 behind the tunnel and a second arm 514 over the top of the tunnel, for example. A linear or other shaped detector array may be used, as well. The tunnel 510 has windows 516 to allow for the passage of an X-ray radiation beam R. Shielding walls 518 surround the source 502, the detector 506, and a portion of the conveying system 508. Concrete is a preferred shielding material for both neutrons and X-rays. If space or other requirements prevent the use of concrete, then a multi-layer shield may be used. For example, polyethylene may be used as an inner layer to shield neutrons and lead or steel may be used as an outer layer to shield X-rays. The outer layer also shields gamma rays emitted by the polyethylene. Openings (not shown) are provided in the shielding walls 518 for the cargo conveyance 504 to be conveyed into and out of the scanning system 500 by the conveyor system 508. The X-ray source 502 may be positioned so that the lower portion of the X-ray radiation beam is parallel or nearly parallel to the top of the conveyor system 508. If the radiation beam R intercepts the conveyor system 508 and the conveyor system 508 comprises a belt or track, a material that causes low attenuation of radiation may be used. If the conveyor system 508 comprises rollers, a gap may be provided among the plurality of rollers, where necessary. A window may also be provided in the structure supporting the conveyor system 508, if necessary. Collimators (not shown) may be provided between the cargo conveyance 504 and the detector 506 to block scattered radiation from reaching the detector 506. The conveyor system 508 may be reversed to examine a portion or the entire cargo conveyance 504 again or to irradiate the cargo conveyance 504 with a different energy distribution, for example. The cargo conveyance 504 may also be irradiated with multiple energies by rapidly cycling between two or more energy levels as the cargo conveyance 504 is being conveyed through the scanning unit 500. The detector 506 is electrically coupled to an image processor block 520, which is coupled to a display 522. The image processor block 520 comprises analog-to-digital conversion and digital processing components, as is known in the art. One or more computers 524 is electrically coupled to and controls the operation of one or more of the X-ray source 500, the detector 506, the conveyor system 508, the image processor 520, and the display 522. The connections between the computer and all the components are not shown, to simplify the Figure. The one or more computers 524 may provide the processing functions of the image processor 520, as well. As shown in FIG. 8, the collimating slot 503a and the X-ray radiation beam R are directed towards the region above the conveyor system 202, to irradiate the cargo conveyance 504. The radiation beam R may diverge over an angle θ. The X-ray source 502 is preferably displaced a sufficient distance from the cargo conveyance 504 so that the beam R intercepts the entire cargo conveyance 504. The angle θ may range from about 30 degrees to about 90 degrees, for example. The configuration of the detector 506 may depend on the shape of a collimated radiation beam. For example, if the collimated radiation beam R is a fan beam, a one-dimensional detector array 504 comprising a single row of detector elements may be provided. If the collimated radiation beam R is a cone beam, the detector array may comprise a two dimensional detector array 506 comprising two or more adjacent rows of detector elements. The detector array 506 may comprise a plurality of modules of detectors, each comprising one or more rows of detector elements supported in a housing. The embodiments described above may also be useful in detecting test materials known to emit neutrons under certain test conditions. For example, the material may undergo photoneutron processes, such as the emission of delayed neutrons. The significantly reduced (or eliminated) neutron generation of this embodiment assists identification and classification of test materials by enabling more accurate determination of the number of neutrons emitted by the test materials. Radiation sources with the low-neutron shielding of the present invention may operate at a single or multiple energy levels. Linear accelerators that may be used to emit radiation at multiple energy levels are described in U.S. Pat. No. 6,366,021 B1, U.S. Pat. No. 4,400,650 and U.S. Pat. No. 4,382,208, which are assigned to the assignee of the present invention and are incorporated by reference, herein. Another linear accelerator that may be used is described in U.S. application Ser. No. 10/745,947, filed on Dec. 24, 2003, which is also assigned to the assignee of the present invention and is incorporated by reference, herein. A Linatron M9 linear accelerator, manufactured by Varian Medical Systems, Inc. of Palo Alto, Calif., may also be used at single or multiple energies. While cargo conveyances are described above, embodiments of the invention may be used to examine other objects, such as luggage, bags, boxes, etc. In addition, the object may be a patient undergoing radiation scanning or radiation therapy. While the charged particles discussed above are electrons and the generated radiation is X-ray radiation, other charged particles, such as protons and deuterons, may be used to generate other types of radiation. For example, gamma ray radiation may be generated by the impact of protons on materials such as lithium, carbon, or sulfur. While the source described above is a linear accelerator, other types of sources may also be used, such as a betatron, cyclotron, or radio frequency quadropole, for example. The embodiments described herein are examples of implementations of the invention. Modifications may be made to these examples without departing from the scope of the invention, which is defined by the claims, below.
	abstract	A processing system includes a particle beam column for generating a particle beam directed to a first processing location; a laser system for generating a laser beam directed to a second processing location located at a distance from the first processing location; and a protector including an actuator and a plate connected to the actuator. The actuator is configured to move the plate between a first position in which it protects a component of the particle beam column from particles released from the object by the laser beam and a second position in which the component of the particle beam column is not protected from particles released from the object by the laser beam.
046876235	summary	CROSS-REFERENCE TO RELATED APPLICATIONS Commonly owned U.S. patent application entitled "Voted Logic Power Circuit and Method of Testing the Same" concurrently filed in the names of Bruce M. Cook and Jerzy Gutman and identified by assignee's docket number 52,263. Commonly owned U.S. patent application entitled "A Voted Logic Power Interface with Tester" concurrently filed in the name of Rober E. Hager and identified by assignee's docket number 52,264. Commonly owned U.S. patent application entitled "Testable, Fault Tolerant Power Interface Circuit for Normally De-Energized Loads", concurrently filed in the name of Robert A. Hager and identified by assignee's docket No. 52,579. BACKGROUND OF THE INVENTION 1. Field of the Invention This invention is directed to systems used to provide automatic responses to abnormal conditions in complex processes such as nuclear reactors, and to apparatus for testing such systems. More particularly, it is directed to means for reliably testing such systems for both normally energized and normally deenergized response devices without interrupting system response to abnormal conditions and despite large variations in energizing voltage. 2. Prior Art Protection systems for complex processes monitor selected process parameters, such as temperatures, pressures and flows, and the status of various components such as whether a valve is open or closed or whether a pump is on or off, and provide automatic responses to measured values of the parameters and to detected status states of the components which require positive intervention to prevent, or to alleviate the effects of, abnormal process conditions. High reliability is an essential requirement for such a system. In order to enhance reliability, it is common practice to provide redundant sensors for each selected parameter and component status. It is also common practice to vote the responses of the redundant sensors, that is to require that a plurality, but not necessarily all, of the sensors, detect the abnormal condition before action is initiated, in order to reduce the probability of a spurious actuation. A nuclear power plant is one example of a complex process in which such a protection system is employed. The protection system in a nuclear power plant performs a plurality of functions. It can shutdown, or trip, the reactor if conditions warrant, or it can perform a number of engineered safeguard functions, such as opening or closing valves and turning on or off pumps or other components. Typically, the trip function involves deenergizing electro-mechanical jacks which normally hold control rods in a position withdrawn from the reactor core so that the rods reenter the core and cause it to go subcritical. The engineered safeguard functions may involve either deenergizing a load device which is normally energized or energizing a device which is normally deenergized. In a typical engineered safeguard function system, four redundant sensors are used to detect the selected parameters and/or status conditions. The response of each sensor is compared with a setpoint value to generate a digital signal which is referred to as a partial actuation signal since an indication from more than one sensor is required to actuate the safety component. The four partial actuation signals for each parameter or status condition are all fed to each of two identical, electrically isolated logic trains. Typically, this is accomplished by applying each partial actuation signal to the coil of a relay having one set of contacts in each logic train. Each logic train independently votes the partial actuation signals, such as two out of four, and generates an actuation signal. The two independently generated actuation signals are then applied to a power interface circuit which requires the presence of both actuation signals to actuate the load device, either a normally energized or normally deenergized component, to initiate the engineered safeguard function. Such a two out of two voting power interface can be disabled by a single failure in one of the two channels. In order to provide tolerance to single failures in a logic train or switching device, the systems described in the related applications referred to above propose the use of two out of three voting power interfaces. Regulations require that the switching devices comprising the power interface be tested periodically. At present, these tests are performed manually with the plant remaining on line. To avoid disrupting plant operation, special test procedures and circuits have been employed to permit testing without changing the energization status of the actuated device associated with the interface under test. In the case of a normally energized load which cannot be deenergized while the plant is in operation, the apparatus and method used are as described in U.S. Pat. No. 3,967,257. This involves connecting a current monitor in series with the switching device under test and connecting in parallel with that combination, a second switching device which is also equipped with a visual current monitor. To perform the test, the second switching device is first "closed" in order to maintain power to the load. The device under test is then exercised while the corresponding current monitor is observed as an indication of its switching state. Normally, deenergized loads which cannot be energized during testing are generally tested by exercising the switching devices using a current which is of sufficient magnitude to be detectable but which is below the actuation current threshold for the actuated device. The prior art systems for testing power interfaces utilize feedback signals which indicate the presence or absence of current in the various circuit legs or they generate analog or digital representations of current magnitude. One problem with test schemes which rely on reading current magnitude is that the current varies as a function of power supply voltage. In the case of a nominal 120 volt DC system, a voltage swing of 50 volts may occur between a low battery condition (approximately 100 VDC) and a full battery or charging condition (approximately 150 VDC). A primary object of the subject invention is to provide a testable voted logic protection system and particulary a power interface for such a system which is operative with either normally energized or normally deenergized loads without interrupting the protection function and without a change in circuit topology. It is another important object of the invention to provide such apparatus which is self-compensating for large variations in power supply voltage. It is still another important object of the invention to provide such apparatus which generates reliable, one bit digital signals in response to test signals. SUMMARY OF THE INVENTION These and other objects are realized by an n out of m voted logic power interface circuit in which m sets of switches are arranged in a plurality of groups of switches with the groups of switches connected in parallel with each other and in series with a voltage source and a load. Each group of switches includes a different selection of n switches connected in series, each from a different one of the m sets of switches. The plurality of groups of switches include all possible combinations of m sets taken n at a time such that with at least n out of the m sets of switches actuated the load device is actuated. Each of the switches is shunted by a resistor to provide a leakage path through each group of switches. However, the impedance of the shunt resistors is several magnitudes greater than that of a closed switch so that the leakage current is insufficient to energize the load. Detectors associated with each group of switches, and responsive to changes in impedance in the group, generate output signals indicative of the state of the switches. Preferably, the detectors generate digital signals having a first value when none of the switches are actuated and a second value when at least one switch in the group is actuated. As used throughout the specification and claims, the term actuated means that the referenced device is in its operated condition. Thus, a normally closed switch is open when actuated, and a normally open switch is closed. Likewise, a normally deenergized load is energized in its actuated state and a normally energized load is deenergized. In order to provide self-compensation for variations in supply voltage, each group of switches is incorporated into a resistance measuring bridge circuit in which the voltage drop across the resistor shunted switches is compared with a reference voltage. Both of these voltages are proportional to the supply voltage so that reliable indications of switch actuation are generated despite even large fluctuations in supply voltage. Thus, in a wheatstone bridge circuit, the voltage drop across the group of resistor shunted switches forming one leg of the bridge is applied to one input of a comparator and the reference voltage generated by the other side of the bridge is applied to the other comparator input. The three reference resistors in the bridge circuit are selected such that the comparator has two discrete outputs; one when none of the switches are actuated and another when at least one of them is actuated. Preferably, the reference resistors are selected such that the reference voltage developed by the bridge and applied to the comparator is about halfway between the voltage drop across the group of resistor shunted switches when no switches are actuated, and that when at least one switch is actuated, for all supply voltages. Where the switches are normally closed devices, a bias voltage is added to the reference voltage to assure reliable switching by the comparator. While suitable for use in other applications, the invention is particularly adapted for use in a protection system for a nuclear power plant where redundant sets of sensors monitor selected reactor parameters and multiple logic trains independently generate a voted logic actuation signal for each set of switches from redundant signals generated by the sensors. A test unit selectively generates test actuation signals and monitors the detectors for preselected patterns of digital signals. By generating fewer actuation signals than are required to actuate the load, the tests can be performed while the protection system remains on line, for both normally energized and normally deenergized loads. Hence, the protection function is not interrupted during test and no changes in circuit topology are required.
042773059	abstract	A device for producing hot plasmas comprising a single turn theta-pinch coil, a fast discharge capacitor bank connected to the coil, a fuel element disposed along the center axis of the coil, a predetermined gas disposed within the theta-pinch coil, and a high power photon, electron or ion beam generator concentrically aligned to the theta-pinch coil. Discharge of the capacitor bank generates a cylindrical plasma sheath within the theta-pinch coil which heats the outer layer of the fuel element to form a fuel element plasma layer. The beam deposits energy in either the cylindrical plasma sheath or the fuel element plasma layer to assist the implosion of the fuel element to produce a hot plasma.
061047732	claims	1. A fuel rod for a nuclear reactor, comprising: a metal cladding tube filled with nuclear fuel and having ends, an outer surface, and a longitudinal axis; a metal seal plug welded to one of the ends of said cladding tube at a transition point, and an annular bead disposed on the outer surface of said cladding tube at the transition point; said annular bead having a cylindrical outer jacket surface extending substantially parallel to the longitudinal axis of said cladding tube; and said annular bead being formed of a material formed of the metal of said cladding tube and the metal of said seal plug. 2. The fuel rod according to claim 1, wherein said seal plug and said cladding tube are formed of the same metal. 3. The fuel rod according to claim 1, wherein said annular bead has at least two humps on said cylindrical outer jacket surface, each extending along said jacket surface substantially parallel to the longitudinal axis. 4. The fuel rod according to claim 1, wherein said annular bead is resolidified from a welding melt. 5. The fuel rod according to claim 1, wherein said cylindrical outer jacket surface of said annular bead is mechanically unmachined. 6. The fuel rod according to claim 1, wherein said cylindrical outer jacket surface of said annular bead has an encompassing depression formed therein in said annular bead containing the material of said cladding tube having the same microscopic structure as in said cladding tube.
	abstract	Processes of reductive decontamination using an agent containing at least two kinds of components, and then decomposing the agent using an apparatus for decomposing at least two kinds of chemical substances in the agent, are employed in chemical decontamination. A catalyst decomposition column in an upstream side of an ion exchange resin column and a hydrogen peroxide injection apparatus in a further upstream side, reduce the amount of waste products caused by a chemical decontaminating agent where a mixed decontaminating agent for a composition trapped in a cation resin column and for a composition trapped in an anion exchange resin are used for the chemical decontaminating agent, selectively decompose the composition trapped in the cation resin column in an inlet side of a cleaning apparatus when radioactive nuclides in the decontaminating agent are cleansed using the cation resin column during decontamination, and decompose both compositions after completion of the decontamination. The chemical decontamination thus selectively decomposes the chemical decontaminating agent, which is a component of the load to the cation resin column. The chemical decontamination moderates corrosion of material by using a chemical decontaminating agent decomposing apparatus capable of decomposing the components trapped by the cation exchange resin and components trapped by an anion exchange resin at the same time.
	claims	1. A wall element consisting of:a first layer of steel;a second layer containing at least some vanadium;a third layer between the first layer and the second layer; andwherein the steel of the first layer consists of:9.0-12.0 wt. % Cr;0.001-2.5 wt. % W;0.001-2.0 wt. % Mo;0.001-0.5 wt. % Si;up to 0.5 wt. % Ti;up to 0.5 wt. % Zr;up to 0.5 wt. % V;up to 0.5 wt. % Nb;up to 0.3 wt. % Ta;up to 0.1 wt. % N;up to 0.3 wt. % C;up to 0.01 wt. % B;the balance being Fe and other elements, wherein the steel includes not greater than 0.15 wt. % of each of these other elements, and wherein the total of these other elements does not exceed 0.35 wt. %. 2. The wall element of claim 1, wherein the second layer has a thickness that is from 0.1% to 50% of the thickness of the first layer and the third layer has a thickness that is from 0.1% to 50% of the thickness of the first layer. 3. The wall element of claim 1, wherein the second layer has a thickness that is from 1% to 5% of the thickness of the first layer and the third layer has a thickness that is from 1% to 5% of the thickness of the first layer. 4. The wall element of claim 1, wherein the second layer is selected from the vanadium alloys V-20Ti, V-10Cr-5Ti, V-15Cr-5Ti, V-4Cr-4Ti, V-4Cr-4Ti NIFS Heats 1 & 2, V-4Cr-4Ti US Heats 832665 & 8923864, and V-4Cr-4Ti Heat CEA-J57. 5. The wall element of claim 4, wherein the second layer is V-4Cr-4Ti. 6. The wall element of claim 4, wherein the second layer consists of:3.0-5.0 wt. % Cr;3.0-5.0 wt. % Ti; andno more than 0.02 wt. % C;with the balance being V and other elements, wherein the vanadium alloy includes not greater than 0.1 wt. % of each of these other elements, and wherein the total of these other elements does not exceed 0.5 wt. %. 7. The wall element of claim 5, wherein the second layer consists of:3.5-4.5 wt. % Cr;3.5-4.5 wt. % Ti;0.04-0.1 wt. % Si;up to 0.02 wt. % O;up to 0.02 wt. % N;up to 0.02 wt. % C;up to 0.02 wt. % Al;up to 0.02 wt. % Fe;up to 0.001 wt. % Cu;up to 0.001 wt. % Mo;up to 0.001 wt. % Nb;up to 0.001 wt. % P;up to 0.001 wt. % S; andno more than 0.0002 wt. % Cl;with the balance being V and other elements, wherein the vanadium alloy includes not greater than 0.001 wt. % of each of these other elements, and wherein the total of these other elements does not exceed 0.01 wt. %. 8. The wall element of claim 1, wherein the second layer consists of:0.001-0.5 wt. % C;the balance being V and other elements, wherein the second layer includes not greater than 0.1 wt. % of each of these other elements, and wherein the total of these other elements does not exceed 0.5 wt. %. 9. The wall element of claim 8, wherein the second layer includes from 0.1 to 0.3 wt. % C in addition to V. 10. A container comprising a wall element of claim 1. 11. The wall element of claim 1, wherein the second layer is at least 90% vanadium. 12. The wall element of claim 1, wherein the third layer is of nickel, nickel alloy, chromium, chromium alloy, zirconium, or zirconium alloy. 13. A container for holding a nuclear fuel comprising:at least one wall element separating a fuel storage volume from an external environment;the wall element having a first layer of steel separated from a second layer containing at least some vanadium by a third layer between the first layer and the second layer;the first layer of the wall contacting the external environment and the second layer contacting and the fuel storage volume; andwherein the steel of the first layer consists of:9.0-12.0 wt. % Cr;0.001-2.5 wt. % W;0.001-2.0 wt. % Mo;0.001-0.5 wt. % Si;up to 0.5 wt. % Ti;up to 0.5 wt. % Zr;up to 0.5 wt. % V;up to 0.5 wt. % Nb;up to 0.3 wt. % Ta;up to 0.1 wt. % N;up to 0.3 wt. % C;up to 0.01 wt. % B;the balance being Fe and other elements, wherein the steel includes not greater than 0.15 wt. % of each of these other elements, and wherein the total of these other elements does not exceed 0.35 wt. %. 14. The container of claim 13, wherein the container has a shape that is defined by one or more continuously connected wall elements to form a vessel. 15. The container of claim 13, wherein the container is shaped as a cylindrical tube, at least one wall element forming a cylindrical wall of the tube and the fuel storage volume being the inside of the tube. 16. The container of claim 13 further comprising:a nuclear material in the container in contact with the second layer. 17. An article, comprising:an amount of nuclear material;a wall element disposed exterior to the nuclear material and separating at least some of the nuclear material from an exterior environment, the wall element consisting of:a first layer of steel in contact with the external environment; anda second layer containing at least some vanadium in contact with the nuclear material; anda third layer between the first layer and the second layer, the third layer inhibiting the transfer of carbon from the steel into the second layerwherein the steel of the first layer consists of:9.0-12.0 wt. % Cr;0.001-2.5 wt. % W;0.001-2.0 wt. % Mo;0.001-0.5 wt. % Si;up to 0.5 wt. % Ti;up to 0.5 wt. % Zr;up to 0.5 wt. % V;up to 0.5 wt. % Nb;up to 0.3 wt. % Ta;up to 0.1 wt. % N;up to 0.3 wt. % C;up to 0.01 wt. % B;the balance being Fe and other elements, wherein the steel includes not greater than 0.15 wt. % of each of these other elements, and wherein the total of these other elements does not exceed 0.35 wt. %. 18. The article of claim 17, wherein the nuclear material includes at least one of U, Th, Am, Np, and Pu. 19. The article of claim 18, wherein the nuclear material and the wall element are mechanically bonded. 20. The article of claim 19, wherein the exterior environment includes molten sodium and the first layer of steel prevents contact between the sodium and the vanadium in the second layer. 21. The container of claim 13, wherein the second layer is at least 90% vanadium. 22. The container of claim 13, wherein the third layer is of nickel, nickel alloy, chromium, chromium alloy, zirconium, or zirconium alloy.
	abstract	A method of servicing a nuclear reactor during a reactor outage is provided. The reactor includes a primary containment vessel and a reactor pressure vessel positioned in the primary containment vessel. The method includes positioning a servicing platform above the reactor pressure vessel and performing predetermined servicing operations on the reactor. The servicing platform includes a frame having a plurality of interconnected beams, a support structure attached to the frame, and a floor attached to a top of the frame. The floor includes a reactor access opening sized to permit access to the reactor pressure vessel. The servicing platform also includes at least one auxiliary platform movably coupled to the frame and extending into the access opening. The at least one auxiliary platform is movable along a perimeter of the access opening of the floor.
060118261	abstract	A steam power station, in particular a nuclear power station, includes a steam conduit leading through a wall and forming a fixed point with the wall for the introduction of forces and moments. A main valve is connected to the steam conduit at the fixed point, without a high-pressure pipe being interposed. Satellite valves, which have smaller nominal widths than the main valve, are fastened to the housing of the main valve, without a high-pressure pipe being interposed. At least one additional valve has a housing fastened to the housing of a satellite valve, without a high-pressure pipe being interposed and without any support.
045432317	claims	1. A method for generating and containing plasma with a multipole plasma pinch comprising: generating plasma; passing current through the plasma in a plurality of discrete channels to form respective z-pinches therein which generate at least one hyperbolic magnetic axis within said plasma, each of said channels being disposed within the plasma and containing a respective set of nested closed magnetic surfaces defining a magnetic axis extending in the direction of current flow, said channels and said at least one hyperbolic axis being surrounded by an additional set of nested closed magnetic surfaces within the plasma; and forming an average magnetic well encompassing substantially all of said plasma while maintaining a safety factor q having an absolute value less than 1 substantially everywhere in the plasma. means for generating plasma and maintaining a safety factor q having an absolute value less than 1 substantially everywhere in the plasma; means for passing current through the plasma in a plurality of discrete z-pinch channels to form respective z-pinches therein which generate at least one hyperbolic magnetic axis within said plasma, each of said channels being disposed within the plasma and containing a respective set of nested closed magnetic surfaces defining a magnetic axis extending in the direction of current flow, said channels and said at least one hyperbolic axis being surrounded by an additional set of nested closed magnetic surface within the plasma; and means for forming an average magnetic well encompassing substantially all of said plasma. 2. The method according to claim 1 wherein said step of forming an average magnetic well includes making the component of magnetic field in the direction of said at least one hyperbolic axis substantially zero in the vicinity of such axis. 3. The method according to either one of claims 1 and 2 wherein said additional set of nested closed magnetic surfaces is bounded internally by a plurality of respective separatrix surfaces surrounding respective channels, said at least one hyperbolic axis lying on respective separatrix surfaces, and wherein said step of forming an average magnetic well includes making the component of magnetic field in the direction of said at least one hyperbolic axis reverse direction within said plasma outside said separatrix surfaces. 4. The method according to claim 3 wherein said well is formed substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 5. The method according to claim 4 wherein said channels are shaped and positioned by means external to the plasma. 6. The method according to claim 3 wherein said channels are shaped and positioned by means external to the plasma. 7. The method according to either one of claims 1 and 2 wherein said channels are shaped and positioned by means external to the plasma. 8. The method according to either one of claims 1 and 2 wherein said well is formed substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 9. The method according to claim 8 wherein said channels are shaped and positioned by means external to the plasma. 10. Apparatus for generating and containing plasma with a multipole plasma pinch comprising: 11. Apparatus according to claim 10 including wall means defining a chamber for containing said plasma, said wall means being spaced from said channels. 12. Apparatus according to claim 11 wherein said chamber and channels are toroidal. 13. Aparatus according to claim 12 wherein said chamber and channels are axisymmetric about the major axis of said chamber. 14. Apparatus according to any one of claims 10 to 13 including means external to the plasma for shaping and positioning said channels. 15. Apparatus according to claim 14 wherein said means for shaping and positioning includes an electrically conducting shell of the shape desired for said plasma. 16. Apparatus according to claim 14 wherein said means for shaping and positioning includes field shaping coils. 17. Apparatus according to any one of claims 10 to 13 wherein said means for passing current includes induction coils that are disposed external to the plasma for inducing said current and are distributed for shaping and positioning said channels. 18. Apparatus according to any one of claims 10 to 13 wherein said means for forming an average magnetic well includes means for making the component of magnetic field in the direction of said at least one hyperbolic axis substantially zero in the vicinity of such axis. 19. Apparatus according to claim 14 wherein said means fo forming an average magnetic well includes means for making the component of magnetic field in the direction of said at least one hyperbolic axis substantially zero in the vicinity of such axis. 20. Apparatus according to any one of claims 10 to 13 wherein said additional set of nested closed magnetic surfaces is bounded internally by a plurality of respective separatrix surfaces surrounding respective channels, said at least one hyperbolic axis lying on respective separatrix surfaces, said apparatus including means for making the component of magnetic field in the direction of said at least one hyperbolic axis reverse direction within said plasma outside said separatrix surfaces. 21. Apparatus according to claim 14 wherein said additional set of nested closed magnetic surfaces is bounded internally by a plurality of respective separatrix surfaces surrounding respective channels, said at least one hyperbolic axis lying on respective separatrix surfaces, said apparatus including means for making the component of magnetic field in the direction of said at least one hyperbolic axis reverse direction within said plasma outside said separatrix surfaces. 22. Apparatus according to claim 18 wherein said additional set of nested closed magnetic surfaces is bounded internally by a plurality of respective separatrix surfaces surrounding respective channels, said at least one hyperbolic axis lying on respective separatrix surfaces, said apparatus including means for making the component of magnetic field in the direction of said at least one hyperbolic axis reverse direction within said plasma outside said separatrix surfaces. 23. Apparatus according to claim 19 wherein said additional set of nested closed magnetic surfaces is bounded internally by a plurality of respective separatrix surfaces surrounding respective channels, said at least one hyperbolic axis lying on respective separatrix surfaces, said apparatus including means for making the component of magnetic field in the direction of said at least one hyperbolic axis reverse direction within said plasma outside said separatrix surfaces. 24. Apparatus according to any one of claims 10 to 13 wherein said means for forming an average magnetic well includes means for forming said well substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 25. Apparatus according to claim 14 wherein said means for forming an average magnetic well includes means for forming said well substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 26. Apparatus according to claim 18 wherein said means for forming an average magnetic well includes means for forming said well substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 27. Apparatus according to claim 20 wherein said means for forming an average magnetic well includes maans for forming said well substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 28. Apparatus according to claim 23 wherein said means for forming an average magnetic well includes means for forming said well substantially at a magnetic surface within said plasma where the safety factor q is substantially zero. 29. A multipole pinch plasma device comprising wall means forming a chamber; and means for generating plasma within said chamber and maintaining a safety factor q having an absolute value less than 1 substantially everywhere within said plasma and for producing an average magnetic well encompassing said plasma, including means for producing current in said plasma in a plurality of discrete z-pinch current channels to form respective z-pinches therein, each of said channels containing a set of nested closed magnetic surfaces defining a magnetic axis in the direction of current flow, said channels being spaced from said wall means and disposed within said plasma to form a hyperbolic magnetic axis therein, said hyperbolic magnetic axis and said current channels being surrounded by a set of nested closed magnetic surfaces.
059189118	abstract	For nozzle replacement, repair and initial installation in ASME pressure vessels, disclosed are nozzles, nozzle assemblies and nozzle repair assemblies, all of which are mechanically attached and mechanically sealed to the vessel, i.e., without any welding to the vessel, and in most embodiments without any welding at all. For nozzle replacement, a part or the entire existing nozzle is removed and a partial or full replacement nozzle or nozzle assembly is mechanically attached and mechanically sealed to the vessel. For nozzle repair, the existing nozzle within the bore is not removed, and a mechanical seal is provided for the existing nozzle which may also be welded or mechanically attached to the repair assembly. Also disclosed are: a corrosion resistant leak path or paths past the nozzle to the exterior of the vessel and beyond the insulation surrounding the vessel such that any leakage can visually be detected; an anti-rotation device or devices which prevent a mechanically attached nozzle part from rotating relative to the vessel; and an assembly which rotatably couples a sleeve threaded to the bore or a clamp device and a nozzle or nozzle body so that the sleeve may be tightened without breaking a weld between the nozzle and the assembly which attaches the nozzle to the sleeve; and a clamp which releasably frictionally engages a nozzle and releasably attaches it to a sleeve or flange that is mechanically attached to the vessel.
	claims	1. A method for measuring a dimension of a pattern using a plurality of scanning electron microscopes (SEMs),comprising the steps of:irradiating and scanning a converged electron beam of a first SEM on a sample on which a pattern is formed;acquiring an image of the pattern formed on the sample by detecting secondary electrons generated from the sample by the irradiation of the converged electron beam, and creating at least one image profile from the acquired image of the pattern acquired by the first SEM;reading filter parameter data from a memory, and using the data for adjusting feature quantities of the image profileto match with corresponding feature quantities of at least one image profile of the pattern obtained by using at least one SEM from the plurality of SEMs that is different from the first SEM; andmeasuring the dimension of the pattern from the adjusted feature quantities of the image profile. 2. The method for measuring a pattern dimension using a plurality of scanning electron microscopes (SEMs) according to claim 1,whereinthe filter parameter is selected to produce matching between the feature quantities of the image profiles obtained by using the first SEM and corresponding feature Quantities of the image profile obtained by using the at least one SEM from the plurality of SMS that is different from the first SEM, so that the differences between corresponding feature quantities are minimized. 3. The method for measuring a pattern dimension using a plurality of scanning electron microscopes (SEMs) according to claim 1,whereina filter parameter by which the corresponding feature quantities of the image profile obtained by the first SEM and the image profile obtained by the at least one SEM from the plurality of SMS that is different from the first SEM are matched, is the filter parameter obtained by comparing pieces of information of the image profiles obtained by imaging the same sample pattern for the same number of times using the first SEM and the at least one SEM from the plurality of SMS that is different from the first SEM. 4. The method for measuring a pattern dimension using a plurality of scanning electron microscopes according to claim 1,further comprising the step of:reading a magnification correction parameter that is stored in the memory in advance and used to match magnifications between the first SEM and the at least one SEM from the plurality of SMS that is different from the first SEM,whereinthe magnification correction parameter is a parameter calculated from image profiles obtained by imaging repeated patterns each having the same shape formed on the sample using first SEM and at least one SEM from the plurality of SMS that is different from the first SEM. 5. The method for measuring a pattern dimension using a plurality of scanning electron microscopes (SEMs) according to claim 1,whereinthe filter parameter by which the corresponding feature quantities of the image profiles are matched is a parameter that considers a difference between a beam shape of the first SEM and a beam shape of the at least one SEM from the plurality of SMS that is different from the first SEM. 6. A method for measuring a dimension of a pattern using a plurality of scanning electron microscopes (SEMs) comprising the steps of:designating a scanning electron microscope to serve as a reference SEM;designating a SEM to be calibrated;calculating a feature quantity of an image profile of a sample pattern, obtained by imaging the sample pattern with the reference SEM;calculating a feature quantity of an image profile of the sample pattern obtained by imaging the sample pattern with the SEM to be calibrated;finding a value of a parameter by which the calculated feature quantity of the image profile of the sample pattern obtained by imaging with the reference SEM is matched with the feature quantity of the image profile of the sample pattern obtained by imaging with the SEM to be calibrated;correcting the sample image obtained by imaging with the SEM to be calibrated, using the parameter value by which the corresponding feature quantities of the images obtained using the reference SEM and the SEM to be calibrated are matched; andfinding the dimension of the sample pattern from the corrected sample image. 7. The method for measuring a pattern dimension using a plurality of scanning electron microscopes (SEMs) according to claim 6,whereinfinding the value of the parameter is performed by the steps of: calculating the feature quantity of an image profile of the sample pattern after the feature quantity of the image profile of the sample pattern is obtained by imaging the sample pattern two or more times with the reference SEM;calculating the feature quantity of the image profile of sample pattern after the sample pattern is imaged for a plurality of times using the SEM to be calibrated; anddetermining the value of the parameter by which the calculated feature quantity of the image profile of the sample pattern imaged with the SEM to be calibrated is matched with the calculated feature quantity of the image profile of the sample pattern imaged with the reference SEM. 8. A plurality of scanning electron microscopes (SEMs) for measuring a dimension of a pattern,comprising:electron beam irradiation means for irradiating and scanning a converged electron beam on a sample on which a pattern is formed;image acquisition means for acquiring an image of the pattern formed on the sample by detecting secondary electrons emitted from the sample under the impact of the converged electron beam using one SEM;a memory for storing at least one filter parameter determined by matching corresponding feature quantities of at least one image profile obtained from the image of the pattern using the one SEM with the feature quantities of at least one image profile obtained using at least one other SEM different from the one SEM;image profile creation means that read the at least one filter parameter stored in the memory and create an image profile from both the at least one read filter parameter and an image of a pattern acquired by the image acquisition means;pattern dimensional measurement means for measuring a dimension of the pattern from the image profile created by the image profile creation means; anddisplay means for displaying results measured by the pattern dimension measurement means. 9. The scanning electron microscopes (SEMs) according to claim 8, furthercomprisinginput means for designating the at least one other SEM. 10. The scanning electron microscopes (SEMs) according to claim 8,further comprisinginput means for designating a filter parameter by which the feature quantity of the image profile is matched with the at least one other SEM, wherein the one other SEM is a reference SEM. 11. The scanning electron microscopes (SEMs) according to claim 8,whereinthe display means display each difference between corresponding feature quantities of the plurality of image profiles obtained from the images acquired using the one SEM and at least one other SEM. 12. An apparatus for correcting a difference among a plurality of scanning electron microscopes (SEMs) for measuring a dimension of a pattern,comprising: designation means for designating both a reference SEM and a SEM to be calibrated using the reference SEM;first memory for storing a feature quantity of an image profile of a sample pattern obtained by imaging the sample pattern with the reference SEM designated by the designation means,second memory for storing a feature quantity of an image profile of the sample pattern obtained by imaging the sample pattern with the SEM to be calibrated designated by the designation means,parameter calculation means for calculating a parameter to match the corresponding feature quantities of the image profiles of the sample pattern stored in the second memory with the feature quantities of the image profiles of the sample pattern stored in the first memory;image correction means for correcting an image of a specimen obtained by imaging with the SEM to be calibrated using the parameter for matching calculated by the parameter calculation means; anddimension measurement means for determining a dimension of the pattern on the specimen from the image of the specimen corrected by the image correction means. 13. An apparatus for correcting a difference between scanning electron microscopes (SEMs) according to claim 12,whereinthe sample pattern is imaged for a plurality of times with the reference SEM,a parameter calculation means that calculate a parameter that comprises a prediction of a variation of the feature quantity of the image profile of the sample pattern caused by a deformation of the sample pattern induced by a plurality of electron beam irradiations from the reference SEM, anda parameter by which the feature quantity of the image profile of the sample pattern of the reference SEM is matched with the feature quantity of the image profile of the sample pattern of the SEM to be calibrated.
	abstract	An operating device pressure Po into a regulator valve and a degree of opening X from the regulator valve are sampled. A speed of change vPo (k) of the Po (k) is calculated from the current Po (k) and the previous Po (k −1). A speed of change vX (k) of the X (k) is calculated from the current X (k) and the previous X (k −1). If both vPo (k) and vX (k) are small, then a weighting value w1 (k) is set to 1 (and set to 0 otherwise), where a fault check indicator Fq (i) for each degree-of-opening category i during a fault check evaluation time interval is calculated from the vPo (k) and vX (k) when w1 (k) is 1 and from a linear approximation formula F1 that indicates the steady-state input/output relationships in the regulator valve when operating properly.
	summary

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